ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 177

DOE Review Release 10/03/2014

Rocky Flats Plant – Occupational Internal 

Dose

ORAUT-TKBS-0011-5 Rev. 03

Effective Date: 09/30/2014

Supersedes: Revision 02

Subject Expert(s): Mutty M. Sharfi, Matthew Arno, David Allen

Document Owner 

Approval: Signature on File Approval Date: 09/16/2014

Mutty M. Sharfi, Document Owner

Concurrence: Signature on File Concurrence Date: 09/16/2014

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 09/16/2014

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 09/18/2014

James P. Griffin, Deputy Project Director

Approval: Signature on File Approval Date: 09/30/2014

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 2 of 177

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/12/2004 00 New document to establish technical basis document for the Rocky 

Flats Plant - occupational internal dose. First approved issue. 

Initiated by Robert Meyer.

12/13/2005 00 PC-1 Approved page change revision initiated to incorporate recent 

direction from NIOSH to include DOL review comments on page 7 in 

Section 5.1. Change made to Table 5.3.1.4.2-2 on page 17 in 

Section 5.3. No sections were deleted. Retraining is not required. 

Initiated by Robert Meyer.

02/01/2007 01 Approved Revision 01 revised to reflect current introductory material. 

Revised in response to Union comments. Change made to Table 5-

9. Constitutes a total rewrite of document. This revision addresses 

Worker Outreach comments as described in CT-0201, CT-0206, and 

CT-0207. Revised to incorporate attribution, per ORAU request. 

The Worker Outreach comments from the June 23, 2004, meeting of 

the United Steelworkers of America Local 8031 and Rocky Flats 

Security Officers Local Union 1 are addressed in Section 5.3.2 and 

5.4.2 regarding internal organ counts; Sections 5.2.1.2, 5.2.2.2, 

5.2.3.2., and 5.2.4.2 regarding particle size and that as many reports 

as possible have been reviewed in the writing of this section. The 

Worker Outreach comments from the June 23, 2004, meeting of 

Colorado State Building and Construction Trades regarding wholebody counting are addressed in Sections 5.3.2 and 5.2.4. 

Incorporates internal and NIOSH formal review comments. This 

revision results no change to the assigned dose and no PER is 

required. Training required: As determined by Task Manager. 

Initiated by Robert Meyer. 

08/17/2007 02 Approved Revision 02 initiated to capture Advisory Board comments 

associated with the June 2007 Advisory Board meeting. 

Incorporation of thorium and internal coworker from ORAUT-OTIB0038 and OCAS-TIB-014. Incorporates formal internal and NIOSH 

review comments. Training required: As determined by the Task 

Manager. Initiated by Mutty M. Sharfi.

09/30/2014 03 Revision initiated to incorporate Advisory Board comments, the 

approval of SEC-00192, and new dose reconstruction approaches in 

assessing tritium, 233U, and recycled uranium. Incorporates 

additional clarifications on current dose reconstruction guidance.

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Mutty M. Sharfi.

TRADEMARK INFORMATION

ABACOS-PlusTM is a trademark of Canberra Industries in the United States and/or other countries.

Microsoft® and Access® are registered trademarks of Microsoft Corporation in the United States 

and/or other countries.

All other trademarks are the property of their respective owners.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 3 of 177

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................. 10

5.1 Introduction ............................................................................................................................ 12

5.1.1 Purpose...................................................................................................................... 13

5.1.2 Scope......................................................................................................................... 13

5.1.3 Special Exposure Cohort............................................................................................ 13

5.2 Source Term .......................................................................................................................... 13

5.2.1 Plutonium ................................................................................................................... 14

5.2.1.1 Isotopic Composition.................................................................................... 14

5.2.1.2 Plutonium Solubility and Particle Size .......................................................... 15

5.2.2 Americium .................................................................................................................. 16

5.2.2.1 Isotopic Composition.................................................................................... 16

5.2.2.2 Americium Solubility and Particle Size ......................................................... 16

5.2.3 Uranium...................................................................................................................... 16

5.2.3.1 Enriched Uranium ........................................................................................ 16

5.2.3.1.1 Isotopic Composition ................................................................. 16

5.2.3.1.2 Enriched Uranium Solubility and Particle Size ........................... 16

5.2.3.2 Depleted Uranium ........................................................................................ 17

5.2.3.2.1 Isotopic Composition ................................................................. 17

5.2.3.2.2 Depleted Uranium Solubility and Particle Size........................... 17

5.2.3.3 Uranium-233 ................................................................................................ 17

5.2.3.4 Recycled Uranium........................................................................................ 18

5.2.4 Thorium ...................................................................................................................... 18

5.2.5 Neptunium.................................................................................................................. 18

5.3 In Vitro ................................................................................................................................... 19

5.3.1 Plutonium Urinalysis ................................................................................................... 19

5.3.1.1 Methods, Units, Isotopes, and Interferences ................................................ 19

5.3.1.2 Plutonium Reporting Levels, Minimum Detectable Activities, and 

Uncertainties................................................................................................ 20

5.3.2 Americium Urinalysis .................................................................................................. 21

5.3.2.1 Methods, Units, Isotopes, and Interferences ................................................ 21

5.3.2.2 Americium Reporting Levels, Minimum Detectable Activities, and 

Uncertainties................................................................................................ 22

5.3.3 Uranium Urinalysis ..................................................................................................... 22

5.3.3.1 Enriched Uranium ........................................................................................ 22

5.3.3.1.1 Methods, Units, Isotopes, and Interferences.............................. 22

5.3.3.1.2 Reporting Levels, Minimum Detectable Activities, and 

Uncertainties ............................................................................. 22

5.3.3.2 Depleted Uranium ........................................................................................ 23

5.3.3.2.1 Methods, Units, Isotopes, and Interferences.............................. 23

5.3.3.2.2 Depleted Uranium Reporting Levels, Minimum Detectable 

Activities, and Uncertainties ...................................................... 23

5.3.4 Gross Alpha Urinalysis ............................................................................................... 24

5.3.4.1 Methods, Units, Isotopes, and Interferences ................................................ 24

5.3.4.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties............ 24

5.3.5 Tritium ........................................................................................................................ 26

5.3.5.1 Pre-1973 Unmonitored Tritium Exposure ..................................................... 26
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 4 of 177

5.3.5.2 1973 Tritium Release Exposure Method ...................................................... 27

5.3.5.3 Post-1973 Unmonitored Tritium Exposure.................................................... 31

5.3.5.4 Reporting Levels, Minimum Detectable Activities, and Uncertainties............ 32

5.4 In Vivo.................................................................................................................................... 32

5.4.1 Americium/Plutonium.................................................................................................. 32

5.4.1.1 Methods, Units, Isotopes, and Interferences ................................................ 32

5.4.1.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties............ 33

5.4.2 Thorium/Depleted Uranium......................................................................................... 35

5.4.2.1 Methods, Units, Isotopes, and Interferences ................................................ 35

5.4.2.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties............ 36

5.5 Other Bioassay Data.............................................................................................................. 36

5.5.1 Wound Count Data..................................................................................................... 36

5.5.2 Nasal Smears and Fecal Samples.............................................................................. 37

5.6 Records and Reports ............................................................................................................. 38

5.6.1 Urinalysis Records and Reports ................................................................................. 38

5.6.2 Interpretation of the Urinalysis Record Card ............................................................... 38

5.6.3 Interpretation of the Health Sciences Data System – Urinalysis Detail Report ............ 38

5.6.4 Interpretation of Other Urinalysis Reports................................................................... 39

5.6.5 Lung Count Records and Reports .............................................................................. 41

5.7 Attributions and Annotations .................................................................................................. 44

References ........................................................................................................................................ 59

Glossary ............................................................................................................................................ 64

ATTACHMENT A, MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT 

ROCKY FLATS.................................................................................................... 68

ATTACHMENT B, MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT 

ROCKY FLATS.................................................................................................... 88

ATTACHMENT C, EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS....................... 106

ATTACHMENT D, INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT .... 138
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 5 of 177

LIST OF TABLES

TABLE TITLE PAGE

5-1 Weight percent and fraction of alpha activity for weapons-grade plutonium ........................... 14

5-2 Weight percent and fraction of alpha activity for ZPPR plutonium .......................................... 14

5-3 Weight percent and fraction of alpha activity for EU ............................................................... 16

5-4 Weight percent and fraction of alpha activity for DU............................................................... 17

5-5 Activity fraction of contaminant in recycled uranium ............................................................... 18

5-6 Median MDA values for plutonium.......................................................................................... 21

5-7 Median MDA values for americium......................................................................................... 22

5-8 Median MDAs for EU .............................................................................................................23

5-9 Median MDAs for DU from 1952 to April 1964........................................................................ 24

5-10 Median MDAs for DU from May 1964 to the present .............................................................. 24

5-11 Median MDAs for gross alpha measurements........................................................................ 25

5-12 Summary of RFP tritium dose estimates ................................................................................ 31

5-13 Summary of MDAs for 241Am.................................................................................................. 34

A-1 Method codes ........................................................................................................................70

A-2 Correlation of method code and analyte................................................................................. 71

A-3 Values of tolerance and reporting levels................................................................................. 71

A-4 Sample volumes for routine 24-hour urine samples................................................................ 76

A-5 Detector background for gas flow proportional counters......................................................... 76

A-6 Median and 95th-percentile blank count rates........................................................................ 77

A-7 Efficiencies of alpha-counting detectors ................................................................................. 78

A-8 Recoveries used in MDA assessments .................................................................................. 78

A-9 Fraction of alphas absorbed in residue................................................................................... 79

A-10 Gross alpha calibration factor................................................................................................. 80

A-11 Values of variables and MDA for plutonium for median conditions ......................................... 81

A-12 Values of variables and MDA for plutonium for extreme conditions ........................................ 81

A-13 MDA for plutonium for one, two or three extreme conditions .................................................. 81

A-14 Values of variables and MDA for EU for median conditions.................................................... 82

A-15 Values of variables and MDA for EU for extreme conditions .................................................. 82

A-16 MDA for EU for one, two, or three extreme conditions............................................................ 83

A-17 Values of variables and MDA for fluorimetric measurements of DU for median and 

extreme conditions.................................................................................................................83

A-18 Values of variables and MDA for americium for median conditions ........................................ 84

A-19 Values of variables and MDA for americium for extreme conditions ....................................... 84

A-20 Values of the MDA for americium for one, two, or three extreme conditions........................... 84

A-21 Values of variables and MDA for gross alpha measurements for median conditions .............. 85

A-22 Values of variables and MDA for gross alpha measurements for extreme conditions............. 85

A-23 Values of the MDA for gross alpha measurements for one, two, or three extreme 

conditions...............................................................................................................................86

B-1 Discontinuity factors...............................................................................................................93

B-2 Calibration factors for the Ortec germanium detector system ................................................. 97

B-3 Calibration factors for the PGT I germanium detector system ................................................ 98

B-4 Calibration factors for the PGT II germanium detector system ............................................. 100

B-5 Values of variables for the PGT organ pipe germanium detector system ............................. 100

B-6 Calibration factors for the PGT organ pipe germanium detector system............................... 100

B-7 Values of variables for the EG&G organ pipe germanium detector system .......................... 101

B-8 Calibration factors for the EG&G organ pipe germanium detector system............................ 101

B-9 MDA conversion factors for values of ppm 241Am ................................................................. 102

B-10 Americium ingrowth in RFP plutonium.................................................................................. 103

B-11 Americium MDA for in vivo lung counts at RFP.................................................................... 104
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 6 of 177

D-1 Summary of uranium urinary excretion rate analyses, 1953 to 1988 .................................... 144

D-2 Summary of plutonium urinary excretion rate analyses, 1952 to 1988.................................. 145

D-3 241Am lung count bioassay data for individualized 239Pu Type S fits...................................... 147

D-4 Derived uranium intake rates, 1953 to 2005......................................................................... 150

D-5 Derived Type M plutonium intake rates, 1952 to 2005.......................................................... 150

D-6 Derived Type S plutonium intake rates, 1952 to 2005 .......................................................... 151

D-7 IMBA-derived uranium intake rates ...................................................................................... 161

D-8 IMBA-derived plutonium/americium intake rates, Type M..................................................... 167

D-9 IMBA-derived plutonium/americium intake rates, Type S ..................................................... 177

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (1) (first 

activity date on the HSDS portion: 10-29-54) ...................................................................... 107

C-2 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (2) (first 

activity date on the HSDS portion 8-19-53) .......................................................................... 108

C-3 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (3) (first 

activity date on the HSDS portion: 1-6-58) .......................................................................... 109

C-4 Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58).................. 110

C-5 Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73).................. 111

C-6 Analytical Report – Bioassay Analysis Data 3-15-93............................................................ 112

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 .......................................................... 113

C-8 Form 1 – Sample Results 1-29-96........................................................................................ 114

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U)............ 115

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu) .............................. 116

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 117

C-12 General Engineering Laboratories, Inc. 6-28-99................................................................... 118

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 119

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 120

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 121

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 122

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 123

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 124

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................. 125

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 126

C-21 Body Counter Results 12-8-81 ............................................................................................. 127

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 128

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 129

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 130

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 131

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 132

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 133

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 134

C-29 ABACOS-Plus 3-6-96........................................................................................................... 135

C-30 ABACOS-Plus 11-15-01....................................................................................................... 136

C-31 ABACOS-Plus 6-14-01......................................................................................................... 137

D-1 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, Type F ................................................................................................................ 151
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 7 of 177

D-2 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type F. ............................................................................................................... 152

D-3 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type F..................................................................................... 152

D-4 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 84thpercentile, Type F ................................................................................................................ 153

D-5 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, Type M................................................................................................................ 153

D-6 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type M................................................................................................................ 154

D-7 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type M.................................................................................... 154

D-8 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 84thpercentile, Type M................................................................................................................ 155

D-9 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 50thpercentile, Type S................................................................................................................ 155

D-10 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 50thpercentile, Type S................................................................................................................ 156

D-11 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 50thpercentile, Type S................................................................................................................ 156

D-12 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 50thpercentile, Type S................................................................................................................ 157

D-13 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type S................................................................................................................ 157

D-14 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 84thpercentile, Type S................................................................................................................ 158

D-15 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 84thpercentile, Type S................................................................................................................ 158

D-16 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 84thpercentile, Type S................................................................................................................ 159

D-17 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 84thpercentile, Type S................................................................................................................ 159

D-18 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, Type S................................................................................................................ 160
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 8 of 177

D-19 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type S .................................................................................... 160

D-20 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 84th-percentile, Type S .................................................................................... 161

D-21 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, Type M................................................................................................................ 162

D-22 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 50thpercentile, Type M................................................................................................................ 162

D-23 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 50thpercentile, Type M................................................................................................................ 163

D-24 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 50thpercentile, Type M................................................................................................................ 163

D-25 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, Type M................................................................................................................ 164

D-26 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 84thpercentile, Type M................................................................................................................ 164

D-27 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 84thpercentile, Type M................................................................................................................ 165

D-28 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 84thpercentile, Type M................................................................................................................ 165

D-29 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 50-percentile, Type M....................................................................................... 166

D-30 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 84th-percentile, Type M.................................................................................... 166

D-31 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, Type S................................................................................................................ 167

D-32 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 50thpercentile, Type S................................................................................................................168

D-33 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 50thpercentile, Type S................................................................................................................ 168

D-34 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 50thpercentile, Type S................................................................................................................ 169

D-35 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, Type S................................................................................................................ 169
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 9 of 177

D-36 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, Type S................................................................................................................ 170

D-37 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 84thpercentile, Type S................................................................................................................ 170

D-38 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 84thpercentile, Type S................................................................................................................ 171

D-39 Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 84th-percentile, Type S .................................................................................... 171

D-40 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, Type S................................................................................................................ 172

D-41 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 50thpercentile, Type S................................................................................................................ 172

D-42 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 84thpercentile, Type S................................................................................................................ 173

D-43 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

50th-percentile, Type S........................................................................................................ 174

D-44 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

50th-percentile, Type S........................................................................................................ 174

D-45 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

50th-percentile, Type S........................................................................................................ 175

D-46 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

84th-percentile, Type S........................................................................................................ 175

D-47 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

84th-percentile, Type S........................................................................................................ 176

D-48 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

84th-percentile, Type S........................................................................................................ 176
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 10 of 177

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

CEDR Comprehensive Epidemiology Data Resource

Ci curie

cm centimeter

cpm counts per minute

CWT chest wall thickness

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DTPA diethylenetriaminepentaacetate

DU depleted uranium

EDTA ethylenediaminetetraacetate

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

g gram

GSD geometric standard deviation

hr hour

HSDS Health Sciences Data System

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kilovolts-electron, 1,000 electron volts

kg kilogram

L liter

L X-ray low-energy X-ray

LLNL Lawrence Livermore National Laboratory

MDA minimum detectable activity

min minute

mL milliliter

MLT minutes live time

mm millimeter

MPLB maximum permissible lung burden

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System

ORAU Oak Ridge Associated Universities
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 11 of 177

pCi picocurie

PER Program Evaluation Report

PGT Princeton Gamma Tech

PHA pulse height analysis

POC probability of causation

ppm parts per million

RFP Rocky Flats Plant

ROI region of interest

s second

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TBP tributyl phosphate

TOPO trioxyl phosphene oxide

TTA thenoyltrifluroacetone

U.S.C. United States Code

WG weapons-grade

wk week

yr year

ZPPR Zero Power Physics Reactor

μCi microcurie

μg microgram

μm micrometer

γ gamma particle

§ section or sections
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 12 of 177

5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historical background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 13 of 177

5.1.1 Purpose

This technical basis document (TBD) discusses Rocky Flats Plant (RFP) internal dosimetry data for 

dose reconstruction and includes guidance for the appropriate use of that information. 

5.1.2 Scope

Workers at RFP had the potential to receive intakes of plutonium, americium, enriched uranium (EU), 

depleted uranium (DU), and tritium, as well as miscellaneous other radionuclides (Daugherty et al. 

2001). Section 5.2 describes the available source term information including isotopic composition, 

solubility, and particle size. Site-specific internal dosimetry information for other radionuclides such as 

thorium, curium, and neptunium is rare or not available.

The primary modes of intake would have been chronic or acute inhalation or through breaks in the 

skin (wounds). The primary bioassay data are the urine data (the activity of the radionuclide of 

interest that is excreted in the urine after an inhalation or wound intake) and the lung count data (the 

activity of the radionuclide present in the lungs after an inhalation intake) [1]. Sections 5.3 and 5.4 

discusses these two data sets in detail including the history, sensitivity, and pertinent nuances of the 

methods and data.

The internal exposure record for a worker consists of records of the bioassay data and reports of 

involvement in incidents, accidents, or special situations. Section 5.6 describes samples of these 

records and reports with explanations of the aspects important to dosimetry.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 5.7.

5.1.3 Special Exposure Cohort

NIOSH has determined that doses to unmonitored RFP workers from neptunium, thorium, and 233U

(and its associated 232U and 228Th progeny) cannot be reconstructed from April 1, 1952, through 

December 31, 1983, inclusive. For this reason, the following class of RFP employees has been 

added to the Special Exposure Cohort (SEC) (NIOSH 2013):

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Rocky Flats Plant in Golden, 

Colorado, from April 1, 1952 through December 31, 1983, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort.

The class includes all workers during the SEC period. Because of the identified dose reconstruction 

infeasibility, all dose reconstructions for monitored workers during the SEC period are considered 

partial dose reconstructions. If monitoring data are available for workers in the SEC, dose is to be 

assigned as appropriate based on that data. However, such dose reconstructions are still considered 

partial dose reconstructions because of the determination that exposure to neptunium, thorium, and 

233U (and its associated 232U and 228Th progeny) during the SEC period cannot be bounded.

5.2 SOURCE TERM

In 1993, the Secretary of Energy formally announced the end of nuclear production at Rocky Flats. 

Remediation was completed at the RFP in late 2005. Coworker intakes should be assigned, when 

applicable, up through 2005. Only environmental intakes should be assigned after 2005.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 14 of 177

5.2.1 Plutonium

5.2.1.1 Isotopic Composition

Three aspects of the isotopic composition of plutonium are important to internal dose reconstruction:

1. The percent by weight of 241Pu, which is needed to calculate the ingrowth of 241Am for the lung 

count data,

2. The fraction of the activity for each alpha-emitting plutonium isotope, which is needed to 

account for the dose from unmeasured isotopes,

3. The ratio of the activity of 241Pu to the alpha activity of the other plutonium isotopes, which is 

needed to calculate the intake of 241Pu from intakes from bioassay data for 239Pu and 240Pu.

For weapons-grade (WG) plutonium, which was present at RFP throughout most of its 1952-to-1989 

production history, the ratio of the activity of 241Pu to the alpha activity of the other plutonium isotopes 

is 5.1 and the 240Pu content is about 6% by weight. Table 5-1 lists the weight percent and fraction of 

alpha activity for each isotope.

Table 5-1. Weight percent and fraction of alpha activity for 

weapons-grade plutonium.

a

Isotope Weight percent

Fraction of 

alpha activityb,c

Pu-238 0.01 0.023

Pu-239 93.79 0.8

Pu-240 5.8 0.18

Pu-241 0.36b –

Pu-242 0.03 Negligible

a. Source: Final Environmental Impact Statement, Rocky Flats Plant Site 

(DOE 1980, Volume 1, Table 2.7.2-2, p. 236). Values are the average for 

RFP plutonium from July 1976 to July 1, 1978. This isotopic composition 

is also typical of plutonium metal processed at RFP to 1990 (James 

1990).

b. The percent by weight of 241Pu for 1959 to 1977 was 0.49, with a range of 

0.35 to 0.65 (RFETS 2002, p. 120).

c. – = not applicable.

The Zero Power Physics Reactor (ZPPR) special project in the mid-1960s involved reactor-grade 

plutonium. The ratio of the activity of 241Pu to the alpha activity of the other plutonium isotopes is 32. 

Table 5-2 lists the weight percent and alpha activity fraction for each isotope. Reports of accidents or 

incidents that involved ZPPR plutonium generally note “ZPPR” or “ZPPR material,” especially on the 

lung count reports [2].

Table 5-2. Weight percent and fraction of alpha activity for 

ZPPR plutonium.

a

Isotope Weight percent

Fraction of 

alpha activityb

Pu-239 87.6 0.7

Pu-240 10.0 0.3

Pu-241 2.4 –

a. These ZPPR values are based on extracted data in a working file from an 

undocumented source.

b. – = not applicable.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 15 of 177

Dose reconstructions should account for the activity of 241Am in the plutonium mixture. The 

concentration of the 241Am is variable depending on the time since the plutonium was purified and 

whether the mixture involved waste or byproduct (separated 241Am) from the purification of aged 

plutonium. Starting in 1969, parts per million of 241Am (ppm 241Am) were measured for the plutonium 

mixture in significant possible inhalation incidents and were generally recorded on lung count reports 

for involved workers. A nominal amount, 100 or 1,000 ppm by mass, of 241Am should be assumed if 

no other data are available. Note that the practice at RFP was to measure the ppm 241Am in a 

representative sample of material that was involved in a possible inhalation incident. If a 

representative sample was not obtained or the origin of the intake was not known, a default value of 

1,000 ppm 241Am was used and was assigned to the date of the intake or to the date of the first 

positive lung count if the date of the intake was not known. The fact that RFP arbitrarily assumed

1,000 ppm should not be the basis for determining the plutonium mixture.

If the plutonium intake for WG plutonium is assessed for 239,240Pu, the activity of 241Am in the intake 

mixture is calculated by multiplying the 239,240Pu activity by [48.2 × ppm 241Am ÷ (1 × 106 – ppm 

241Am)]. For ZPPR plutonium, the 239,240Pu activity is multiplied by [44.6 × ppm 241Am ÷ (1 × 106 – ppm 

241Am)] to obtain the activity of 241Am in the intake mixture [3].

5.2.1.2 Plutonium Solubility and Particle Size

Most plutonium in metalworking operations and fire incidents was insoluble (i.e., type S). Exceptions 

such as plutonium metal in solvents such as carbon tetrachloride can be assumed to be more soluble 

(type M) if this is what the data show or if it is more favorable to claimants to do so [4].

The plutonium fire on October 15, 1965, in Buildings 776 and 777, is a special case. The plutonium, 

which was strongly retained in the lungs of exposed workers with relatively low transfer to the urine, 

exhibited highly insoluble (type Super S) characteristics [5].

Plutonium in chemical processing operations can be either soluble (type M), insoluble (type S), or a 

mixture of solubilities. Dose reconstructors should select the material type that is most favorable to 

the claimant [6]. Lung count data in conjunction with urine data can help to determine absorption 

type.

In general, particle size and distributions are not available for work areas or incidents at RFP. 

Therefore, dose reconstructions should use the default value of 5-μm activity median aerodynamic 

diameter (AMAD) (NIOSH 2002).

One exception is the plutonium fire on October 15, 1965, in Buildings 776 and 777 (Dow 1965a,b), for 

which Mann and Kirchner (1967) measured a mass median diameter of 0.3 μm (1-μm AMAD) with a 

geometric deviation of 1.83. An approach that is favorable to claimants is to assume 1-μm AMAD for 

all plutonium fires unless the qualifying cancer involves the tissues of the extrathoracic regions [7].

The 1-μm particle adjustment for RFP plutonium fires should only be applied for energy employees 

who were involved with a known intake from a plutonium fire (or any time dose reconstructors deem 

use of a 1-μm AMAD particle size appropriate) (NIOSH 2002). This can be from involvement with the 

plutonium fire itself, including being in the building or area and exposed to smoke or airborne activity 

from the fire as well as involvement in cleanup activities immediately after the fire.

The application of the 1-μm particle size adjustment only applies to individuals who were involved in a 

fire (i.e., operators and firefighters) and the individuals who performed the immediate clean up of the 

incident. Once that is accomplished, it is assumed that the particle size reverts back to the default 

5-μm AMAD. When applicable, the adjustment factor is applied only to the dose associated with the 
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 16 of 177

intake that is directly from the fire and cleanup. The 1-μm particle size adjustment typically applies for 

a short period (i.e., days, weeks, or a few months).

The use of the 1-μm particle adjustment for RFP plutonium fires is specific to the intake being 

assessed. If an earlier or later intake is assessed that is not associated with a plutonium fire, the 

1-μm particle adjustment factor does not apply.

5.2.2 Americium

5.2.2.1 Isotopic Composition

For the NIOSH Dose Reconstruction Project, the measured americium is 241Am [8].

5.2.2.2 Americium Solubility and Particle Size

Americium was present in two forms at RFP, as a purified byproduct of plutonium recovery and as 

atoms that are formed by the nuclear transformation of 241Pu and embedded in the matrix of the 

plutonium particle. As a purified byproduct, International Commission on Radiological Protection 

(ICRP) Publication 68 specifies americium inhalation absorption as type M (ICRP 1994a, p. 83). For 

embedded atoms in the matrix of an inhaled plutonium particle, dose reconstructors should use the 

solubility classification for the plutonium particle in Section 5.2.1.2 (ICRP 1994b, p. 79).

Dose reconstructors should use the default 5-μm AMAD particle size (NIOSH 2002) except for fire

incidents, in which a 1-μm AMAD should be assumed for consistency with Section 5.2.1.2 above.

5.2.3 Uranium

5.2.3.1 Enriched Uranium

5.2.3.1.1 Isotopic Composition

Production at RFP involved EU from 1952 to 1963. Table 5-3 lists the weight percent and fraction of 

alpha activity for each isotope.

Table 5-3. Weight percent and fraction of alpha activity for EU.

a

Isotope Weight percent

Fraction of 

alpha activity

U-234 1 0.97

U-235 93 0.031

U-236 0.39 0.0039

U-238 5.4 0.00028

a. Source: DOE (1980, Volume 1, Table 2.7.2-4, p. 238).

5.2.3.1.2 Enriched Uranium Solubility and Particle Size

Operations for EU paralleled those for plutonium and included chemical processing and metalworking. 

Compounds of uranium are generally more soluble than those of plutonium, and solubility 

classification is uncertain. The ICRP assigns UO2(NO3)2 (uranyl nitrate) to inhalation type F; UO3

(yellow cake), UF4, and UCl4 to inhalation type M; and UO2 and U3O8 to inhalation type S (ICRP 1979, 

1994b,c). All of these compounds were involved in the recovery and recycling processes for EU in 

Building 881 (RFETS 2000a).

In many cases, the compound of uranium in an intake was not identified. Dose reconstructors should 

use the solubility classification that is most favorable to claimants.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 17 of 177

If site-specific data for particle size of uranium are not available, dose reconstructors should use the 

default particle size value of 5-μm AMAD (NIOSH 2002).

5.2.3.2 Depleted Uranium

5.2.3.2.1 Isotopic Composition

DU was present at RFP throughout its production history. Uranium-238 accounts for the majority of 

DU internal dose, but the total uranium alpha activity should be included in the dose reconstruction 

(see Table 5-4).

Table 5-4. Weight percent and fraction of alpha activity for DU.

a

Isotope Weight percent

Fraction of 

alpha activity

U-234 0.00058 0.097

U-235 0.23 0.013

U-238 99.77 0.89

a. These values are derived from data in DOE (1980, Volume 1, Table 

2.7.2-4, p. 2-38).

5.2.3.2.2 Depleted Uranium Solubility and Particle Size

Operations with DU involved metalworking including casting, forming, and melting with what was 

probably UO3 and U3O8 (RFETS 2000a). The solubility classification is ambiguous, falling somewhere 

between type S and type M (RFETS 1998a, Section 6.1, p. 165; HPS 1995; Lawrence 1984). Dose 

reconstructors should use the solubility classification that is most favorable to claimants.

If site-specific data for particle size of uranium are not available, dose reconstructors should use the 

default particle size value of 5-μm AMAD (NIOSH 2002).

5.2.3.3 Uranium-233

Operations with 233U (thorium strikes) occurred between 1964 and 1983 (Moment et al., 1999). The 

process included the following steps:

1. Material received as nitrate solution,

2. Thorium strike (thorium fluoride precipitation > peroxide precipitation > UO4 cake),

3. Conversion (UO4 > UO3 > UO2 > UF4),

4. Reduction to metal and casting into an ingot,

5. Rolling ingot into a sheet and producing part blanks from the sheet,

6. Machining, and

7. Sampling.

In the beginning of operations, the first two steps were performed in Building 71 (later called 771). 

Intermediate steps (conversion to UF4, reduction to metal, and casting) were performed in Building 81 

(later called 881). The ingot was rolled and formed into parts in Building 83 (later called 883) and then 

transferred back to Building 81 for final machining. Finished parts were sent to Building 77 (later 

called 777) where they were assembled and shipped. By the mid-1970s, the intermediate steps in 

Building 881 shifted to the research and development areas of Building 771.

Because of data issues and limitations, no specific methods to bound doses from 233U and 232U have 

been determined. Therefore, doses to unmonitored RFP workers from neptunium, thorium, and 233U 

(and its associated 232U and 228Th contaminants) cannot be reconstructed.
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5.2.3.4 Recycled Uranium

For all DOE uranium after 1952, this analysis assumed the possibility that uranium from refineries was 

recycled uranium or contained recycled uranium. Table 5-5 provides the activity fractions that should 

be applied to all uranium intakes after 1952 (NIOSH 2011).

Table 5-5. Activity fraction of contaminant in recycled uranium.

Recycled uranium contaminant Pu-239 Np-237 Tc-99 Th-232 Th-228

Activity fraction of contaminant in uranium 0.00246 0.00182 0.379 2.73E-06 2.73E-06

NOTE: If plutonium intakes are assigned through bioassay or coworker, it should not be assigned as 

part of recycled uranium, as this would result in a double assignment of plutonium. However, the rest 

of the contaminants should still be assigned.

5.2.4 Thorium

Thorium was present at RFP facilities from the beginning of operations in 1952 at least through 1975; 

quantities varied from 0 or gram quantities to 238 kg in any particular month at the site (ChemRisk 

1992; Ulsh 2008; Author unknown 1976). The site used thorium in various ways including:

• Fabrication of metal parts from natural thorium or thorium alloys,

• Use of oxide (“thoria”) as a mold-coating compound,

• In compounds for numerous analytical procedures and research and development programs,

• As a substitute for uranium or plutonium components in various research and development 

activities and programs, and

• The removal of 228Th (thorium strike) performed during 233U processing.

While the consensus of the contributors and authors of the thorium reference documents was that the 

quantities and concentrations of thorium on the site over the years at RFP were minimal, there was 

the potential for thorium exposures to certain populations of workers. The available documentation 

supports the existence of thorium on site in the early 1950s through the development of internal and 

external thorium-monitoring processes (Dow 1953, 1956, 1958).

Because of data issues and limitations, no specific methods to bound doses from thorium have been 

determined. Therefore, NIOSH has determined that unmonitored thorium doses at RFP cannot be 

reconstructed.

5.2.5 Neptunium

Neptunium processing at the RFP included preparation of pure neptunium oxide, metal and metal 

alloys, and the recovery of 237Np from a variety of residues (RFETS 1981). Processes included 

dissolution, anion exchange, precipitation, filtration, calcination, conversion to fluoride, and reduction 

to metal. Fabrication steps such as casting and rolling were also sometimes performed for the 

production of high-purity metal shapes and foils. Neptunium was recovered from residual materials 

including sand, slag, crucibles, casting skulls, and various alloys (with plutonium, tin, uranium, and 

zirconium).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 19 of 177

Because of data issues and limitations, no specific methods to bound doses from neptunium have 

been determined. Therefore, NIOSH has determined that unmonitored neptunium doses at the RFP 

cannot be reconstructed. 

5.3 IN VITRO

5.3.1 Plutonium Urinalysis

5.3.1.1 Methods, Units, Isotopes, and Interferences

Through 1989, the units of the results are dpm/24-hr excretion period (dpm/24-hr sample). After 

1989, the units of the results are dpm/sample regardless of the sample volume or excretion period. 

Spot urine samples for plutonium were rarely requested and were usually associated with a significant 

incident, especially an incident with follow-up DTPA chelation. Assume a 24-hour excretion period 

unless the record indicates that the actual excretion period was different [9].

Through 1977, samples were counted using an air proportional detector system that did not have 

sufficient resolution to separate the alpha energies for the plutonium alpha-emitting isotopes. Starting 

in 1973, an alpha pulse height analysis (PHA) system with surface barrier detectors was phased in 

and had completely replaced the air proportional detector system by 1978. The plutonium urine 

results from the air proportional detector system included activity from 238Pu, 239Pu, and 240Pu. 

Plutonium urine results for samples counted by the PHA system included only 239Pu and 240Pu results. 

Intake assessments are simpler and more favorable to claimants if dose reconstructors assume 239Pu 

and 240Pu for all plutonium urine results unless the worker was involved in a special situation involving 

pure 238Pu. If the intake is assessed using 239Pu and 240Pu data, the 238Pu component of the intake is 

obtained by multiplying the 239Pu and 240Pu intake by 0.0235 [10].

Interferences were probably in the period from 1952 to 1962 because of a lack of specificity of the 

chemical procedure to isolate only the plutonium in the extract. Plutonium results probably included

some americium and thorium activity. In addition, for gross alpha analyses that were assigned to 

plutonium through 1973, the result could have included some contribution from uranium. However, it 

is favorable to claimants to disregard such interferences and take the plutonium results at face value 

unless a value can be determined to be an outlier [11].

From 1963 to 1977, the ion exchange method significantly reduced interferences from americium, 

uranium, and thorium. As the PHA system was phased in starting in 1973, the possibility of 

interferences was further reduced. After 1977, these interferences were not a significant issue for 

plutonium urine results because all samples were counted on the PHA system [12].

Another source of interference was contamination of the tracer (236Pu or 242Pu) by the analyte isotopes 

239Pu and 240Pu, which was an infrequent occurrence [13].

EDTA or DTPA chelation treatments cause enhanced excretion of plutonium in the urine. Urine data 

from within 90 days of a chelation injection have historically been excluded from calculations of 

intakes or depositions of plutonium. Information in the medical or dosimetry records should allow 

dose reconstructors to discern chelation treatments, which generally followed a significant and 

documented incident. In the urine data reports for the Health Sciences Data System (HSDS), urine 

data that was affected by chelation were flagged with a code 1. Code 1 was also used to flag urine 

data that did not pass quality standards. Dose reconstructors should be wary of any urine result 

flagged with a code 1 and in general should not use these data in dose reconstruction [14].
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5.3.1.2 Plutonium Reporting Levels, Minimum Detectable Activities, and Uncertainties

The minimum reporting level for plutonium through 1961 was 0.88 dpm/24-hr sample (this was 10% of 

the RFP tolerance level). For 1962 to April 6, 1970, the minimum reporting level was 0.2 dpm/24-hr

sample. Results less than the reporting level were reported as 0.00 dpm/24-hr sample on computergenerated reports, such as the HSDS (see Attachment C, Figures C-3 and C-4) or background (or 

some abbreviation; e.g., BK) when manually recorded on the Urinalysis Record Card (see Figure C3). For some workers, results initially reported as background on the Urinalysis Record Card were 

superseded by the report of the actual result in reports of the HSDS, if the actual result was 

≥0.00 dpm/24-hr sample. After April 6, 1970, all results ≥0.00 dpm/24-hr sample were reported. 

Negative results were reported as 0.00 dpm/24-hr sample through 1989. After 1989, the actual 

negative value was reported. Starting in approximately 1990, urine results were not normalized to a 

24-hr sample. Instead, the results are dpm/sample regardless of the sample volume [15].

The minimum detectable activity (MDA) for plutonium is presented here for the median conditions. By 

definition of the median value, half of the sample-specific MDAs are lower than the median value, and 

half are higher. In most cases dose reconstructors are not likely to have sufficient data to determine 

the sample-specific MDA, so the median values should be used.

Table 5-6 lists the MDA values for plutonium. The values for 1952 to 1977 are based on examination 

of urinalysis data logs for 1952 to 1971 (see Attachment A). The MDA value for 1971 was 

extrapolated through 1977. The MDA value for 1978 to 1989 is based on matrix blank data (RFETS 

1992) for the routine plutonium urinalysis program for August 1, 1990, to September 27, 1991, using 

blank values with a sample-specific recovery in the range of 0.1 to 1.1 dpm/24-hr sample. This range 

of recoveries mimics the range from 1978 to 1989 for a valid analysis of routine samples. For 1990 to 

1992, the blank values with a sample-specific recovery in the range from 0.35 to 1.1 dpm/24-hr 

sample were used to determine the MDA value. For 1993 to the present, the value of the MDA is 

equal to the sample-specific MDA of 0.02 dpm/sample that was contractually required in the Rocky 

Flats Environmental Technology Site bioassay statement of work (RFETS 1998b) for any laboratory 

that processed the sample, although the required MDA was not consistently achieved by the onsite 

laboratory [16]. Note that the value of the sample-specific MDA is included in the urinalysis data 

reports starting in 1990.

Some urine samples could have been processed by an offsite commercial laboratory before 1993. 

The reports for those samples might have the sample-specific MDAs. If these are not available, the 

MDA in Table 5-6 should be used [17].

Some periods contain transitions that improved the detection of plutonium. For example, from 1964 to 

1977, electrodeposition of the plutonium replaced evaporation of the extract on the planchet. In 

addition, starting in 1973 with four detectors, plutonium samples were processed with an internal 

standard and were counted on a PHA system to establish the sample-specific recovery. The count 

time was also increased to 720 minutes. Because of the difficulty of determining which improvements 

apply to each sample, the MDAs in Table 5-6 do not account for the improvement until the transition

was completed for all samples (i.e., the MDAs are favorable to claimants).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 21 of 177

Table 5-6. Median MDA values for plutonium.

a,b

Period dpm/24-hr sample

1952–1953 0.57c

1954–1962 0.51c

1963 0.44

1964–1977 0.54

1978–1989 0.24

1990–1992 0.24

1993–present 0.02

a. The unit of the MDA values starting in 1990 is dpm/sample.

b. Sample-specific MDA values, if found in the record starting in 1990, 

should be used instead of the generic MDA values in this table.

c. Note that these values of MDA are lower than the reporting level of 

0.88 dpm/24-hr sample used at RFP through 1961. Many urine 

results in this period were rereported with the actual value if greater 

than zero. For those rereported results, these MDA values apply 

instead of the original reporting level.

The uncertainty of the result was not quantified and reported in the record until approximately 1980. 

The reported value was the 2-sigma standard error and included only uncertainties of counting 

statistics that were adjusted by the sample-specific recovery. Starting in approximately 1986, 

contributions from other sources of uncertainty were included, and the reported value was the 1-sigma 

standard error [18]. To estimate the uncertainty for results without a reported uncertainty, a 

reasonable approach is to divide the median MDA value by 3.3, where 3.3 is the sum of kα and kβ, and 

kα = kβ = 1.645 (see Attachment A).

5.3.2 Americium Urinalysis

5.3.2.1 Methods, Units, Isotopes, and Interferences

Attachment A describes the methods through 1971. After 1971, the method for 241Am paralleled that 

for plutonium.

The units of the results are dpm/24-hr excretion period through 1989. After 1989, the units of the 

results are dpm/sample regardless of the sample volume or excretion period [19].

The main interference is thorium, specifically 228Th, which has two alphas with energies similar to 

those of 241Am and has chemical properties similar to those of americium. If the chemical extraction 

procedure for americium was not run precisely, thorium would be eluted from the ion exchange 

column with the americium. When the extract was counted, even with the PHA system, the 228Th 

could not be distinguished from the 241Am [20].

The plutonium-to-americium alpha activity ratio (239,240Pu dpm/24-hr sample divided by 241Am 

dpm/24-hr sample) for paired plutonium and americium urine results provides a credibility check. An 

alpha activity ratio less than 2 (corresponding to a parts-per-million value for 241Am of 10,000 or 

greater) is not credible unless the worker was involved with (1) separated 241Am (Line 1 in 

Building 771), (2) the molten salt process in Building 776, (3) research and development projects 

involving pure americium, (4) material from the ZPPR project, or (5) waste identified for those 

operations [21].

Dose reconstructors should use the plutonium urine data instead of the 241Am urine data to assess 

intakes of WG plutonium [22]. The intake of the 241Am is then calculated from the value of the initial 

parts per million of 241Am measured or assumed for the plutonium mixture involved in the intake.
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5.3.2.2 Americium Reporting Levels, Minimum Detectable Activities, and Uncertainties

The reporting levels for americium were ≥0.24 dpm/24 hr in 1963, ≥0.2 dpm/24 hr from 1964 to 1967, 

and ≥0.30 dpm/24 hr from 1968 to 1971. Results less than the reporting level were reported as zero 

or background (or some abbreviation; e.g., BK). The reporting practice for the period from 1972 to 

1976 has not been determined. Until it is determined, dose reconstructors should assume that the 

reporting level for 1968 to 1971 was continued through 1976 [23]. Starting in 1977, all results 

≥0.00 dpm/24-hr sample were reported. Negative results were reported as zero through 1989. After 

1989, the actual negative value was reported. As for plutonium, urine results were not normalized to 

a 24-hour sample starting in about 1990. Instead, the results are dpm/sample, regardless of the 

sample volume [24].

The MDAs for americium (Table 5-7) were determined as described for plutonium (see Section 5.3.1.2 

and Attachment A), with the difference that the americium analyses started in 1963.

Table 5-7. Median MDA values for americium.

a,b

Period dpm/24-hr sample

1963 0.44

1964–1965c 0.55

1965–1970c 0.46

1971–1977 0.76

1978–1989 0.31

1990–1992 0.3

1993–present 0.02

a. The unit of the MDA values starting in 1990 is dpm/sample.

b. Sample-specific MDA values, if found in the record starting in 1990, 

should be used instead of the generic MDA values in this table.

c. In overlapping years the more favorable MDAs should be assumed.

The discussions of MDA and uncertainty for plutonium urinalysis in Section 5.3.1.2 apply to americium 

urinalysis.

5.3.3 Uranium Urinalysis

5.3.3.1 Enriched Uranium

5.3.3.1.1 Methods, Units, Isotopes, and Interferences

The units of the results are dpm/24-hr excretion period for the entire period. Because urine samples 

analyzed for EU were counted with the air proportional detectors, all of the alpha-emitting isotopes of 

uranium are included in the result. Site-specific information about possible interferences that might 

have occurred for the urinalysis methods for EU is not available. It is favorable to claimants to 

assume that the result is all EU [25].

5.3.3.1.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Table 5-8 lists the MDAs for EU. The reporting level for EU through 1963 was ≥8.8 dpm/24-hr sample

(10% of the RFP tolerance level). From 1964 to 1971, the minimum reporting level ranged from 20 to 

28 dpm/24-hr sample depending on the volume of the sample as observed from the urinalysis data 

logs for that period. Results less than the reporting level were reported as zero or background (or 

some abbreviation; e.g., BK). It is undetermined when urinalysis for EU was stopped at RFP, 

although the stoppage probably occurred in the early 1970s [26].
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Table 5-8. Median MDAs for EU.

Period dpm/24-hr sample

1952–1953 14

1954–1959 13

1960–1963 9.4

1964–1969 31

1970–1971 25

The MDAs for EU were determined as described for plutonium (see Section 5.3.1.2 and 

Attachment A).

Uncertainties for the EU urine results have not been quantified or reported. To estimate the 

uncertainty for results without a reported uncertainty, a reasonable approach is to divide the median 

MDA value by 3.3, where 3.3 is the sum of kα and kβ, and kα = kβ = 1.645 (see Attachment A).

5.3.3.2 Depleted Uranium

5.3.3.2.1 Methods, Units, Isotopes, and Interferences

Attachment A describes the uranium urinalysis methods through 1971. From 1972 to 1979, DU 

samples were chemically processed with the uranium-specific trioxyl phosphene oxide (TOPO) 

extraction procedure, and the electrodeposited extract was counted on the gas flow proportional 

counter. From 1980 to 1997, DU samples were processed with a tracer (232U or 236U) by ion 

exchange and alpha-counted with the alpha spectrometry system with surface barrier detectors in 

vacuum. The starting year of use of the tracer has not been determined. From 1997 to the present, 

DU samples were processed at an offsite commercial laboratory according to provisions of the 

bioassay statement of work (RFETS 1998b).

The units for 1952 to April 1964 were micrograms of uranium per 24-hour excretion period. The mass 

measurement was for all the isotopes of uranium. From May 1964 to 1989, the units were dpm/24-hr

sample. After 1989, the units of the results were dpm/sample, regardless of the sample volume or 

excretion period [27].

The urine data logs through 1971 do not identify the involved isotopes. However, it is reasonable to 

assume that all the alpha-emitting uranium isotopes were included in the air proportional detector 

measurements. For the 1980s, 238U contributes 89% of the alpha activity. Therefore, the logs have 

not been reviewed to determine the other uranium isotopes. Rather, it is favorable to claimants to 

assume that the reported urine result pertains only to 238U and to determine additional intakes for the 

other uranium isotopes [28]. In the 1990s, the urine data reports include the results separately for 

234U, 235U, and 238U.

The major interference is the contribution from natural uranium, which is ubiquitous, sometimes in 

concentrated pockets, in the terrain near RFP. No adjustments have been made to the reported DU 

urine results for this background, which was highly variable.

5.3.3.2.2 Depleted Uranium Reporting Levels, Minimum Detectable Activities, and 

Uncertainties

The minimum reporting level for DU through April 1964 was 5.8 μg/24-hr sample (10% of the 

tolerance level). From May 1964 to 1971, the minimum reporting level was the same as that for EU 

(20 to 28 dpm/24-hr sample depending on the volume of the sample). The reporting level for 1972 to 

1979 (TOPO procedure) has not been determined. An approach that is favorable to claimants is to 

use the reporting level for 1964 to 1971 [29]. In the 1980s, all results ≥0.00 dpm/24-hr sample were 
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reported. Negative values were reported as 0.00 dpm/24-hr sample. In the 1990s and after, all actual 

results, including negative values, were reported.

The MDAs for DU for fluorometric measurements were determined as described in Attachment A. 

Median MDAs for DU from 1952 to April 1964 are listed in Table 5-9. For alpha-counting methods, 

the MDAs in the period from April 1964 to 1971 are the same as those for EU in Table 5-7. The MDA 

value for 1972 to 1979 was extrapolated from the value for the previous period. The MDAs for 1980 

to the present were derived in the same manner as that for plutonium but are based on 238U. 

Table 5-10 lists median MDAs for DU from May 1964 to the present.

Table 5-9. Median MDAs for DU from 1952 to April 1964.

Period dpm/24-hr sample

1952–1955a 31

1955–1959a 12

1960–04/1964 11

a. In overlapping years the more favorable MDAs should be assumed.

Table 5-10. Median MDAs for DU from May 1964 to the 

present.

a,b

Period dpm/24-hr sample

05/1964–1969 31

1970–1971 25

1972–1979 25c

1980–1989 0.56

1990–1992 0.4

1993–present 0.1

a. The MDA value unit starting in 1990 is dpm/sample.

b. Sample-specific MDA values, if found in the record starting in 1990, 

should be used instead of the generic MDA values in this table.

c. Actual practice is unknown; assume continuation of earlier practice.

The discussion of the uncertainty for plutonium in Section 5.3.1.2 applies to DU.

5.3.4 Gross Alpha Urinalysis

5.3.4.1 Methods, Units, Isotopes, and Interferences

Gross alpha measurement is a nonspecific analysis that was used for workers who were potentially 

exposed to both uranium and plutonium in the same monitoring period. Workers who were potentially 

exposed to other alpha-emitting radionuclides, such as neptunium and curium, might also have been 

monitored for gross alpha. Urinalysis methods are discussed in Attachment A. The gross alpha 

method was discontinued in the early 1970s, probably in 1973 [30]. The results are reported as 

dpm/24-hr sample of either EU (the default analyte through 1963) or plutonium (the default analyte 

after 1963). Interferences are likely, because the methods were nonspecific. The analyzed isotopes 

were all of the alpha-emitting isotopes of the analyte.

5.3.4.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

The reporting level for gross alpha through 1963 was ≥8.8 dpm/24-hr sample (10% of the RFP 

tolerance level for EU). After 1963, the reporting level was ≥0.9 dpm/24-hr sample and credited to 

plutonium. (Gross alpha data are probably coded as G in the urine data reports [31].)

Samples with results ≥0.9 dpm/24-hr sample were typically but not always counted using a PHA 

system to determine whether to credit the result to EU, to plutonium, or to a portion to both. The 

default condition through 1963 was to credit the result to EU unless the PHA count indicated 
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otherwise. After 1963 (when EU operations were phased out), the default condition was to credit the 

result to plutonium. In either case, the results should be considered to be upper bounds because of 

the nonspecificity of the analysis [32].

The MDAs for gross alpha in Table 5-11 were determined as described in Attachment A.

Table 5-11. Median MDAs for gross alpha measurements.

Period dpm/24-hr sample

1952 1

1953 0.88

1954–1959 0.79

1960–1962 0.55

1963 0.55

1964–1971 0.69

Uncertainties for the gross alpha urine results have not been quantified or reported. To estimate the 

uncertainty for results without a reported uncertainty, a reasonable approach is to divide the median 

MDA value by 3.3, where 3.3 is the sum of kα and kβ and kα = kβ = 1.645 (see Attachment A). This 

uncertainty does not include the effect of interferences, which is a significant issue for a nonspecific 

analysis like gross alpha measurement [33].
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5.3.5 Tritium

5.3.5.1 Pre-1973 Unmonitored Tritium Exposure

A pre-1973 tritium exposure method was developed based on measurement results provided in a 

Rocky Flats Area Office (RFAO) report issued subsequent to a tritium release in one of the Rocky 

Flats production buildings on August 30, 1974 (AEC ca. 1974). The information contained in this 

report includes measurement data (i.e., results from air samples, surface contamination surveys, and 

bioassay) from the production area where the release occurred as well as comparison data from other 

areas prior to, during, and after the release. Several factors support the use of these data as 

surrogates for bounding the tritium environment at Rocky Flats prior to 1973:

1. Background tritium levels immediately prior to the incident described in the RFAO 

report, although undoubtedly elevated since the more significant 1973 release, were 

well below dosimetrically-significant values and can be considered as fairly 

representative of typical background levels for this analysis. The background tritium 

levels monitored in the months prior to the 1974 incident are consistent with internal 

radiation doses from tritium of well under 1 mrem annually. They are dosimetrically 

insignificant in this sense.

2. The quantity of tritium released (1.5 Ci) was significantly less than that released in 

1973, and is probably more typical of potential undocumented releases in work areas –

particularly those resulting from opening contaminated shipping containers.

The 1974 1.5-Ci tritium release is the only documented release from a shipping 

container in the Rocky Flats workplace. It is taken to be typical since there are no 

other such documented releases to use in forming the model. There is documented 

concern about tritium releases, as shown in the following quote from the ChemRisk 

report (ChemRisk 1994, pdf p. 38):

As early as 1962, Rocky Flats maintained instruments for detection of tritium gas in 

particular work areas of the plant because operations have sometimes resulted in the 

storage of tritium containers.

The instruments available to Rocky Flats at that time were only semi-quantitative for 

indicating the presence of tritium; NIOSH has captured no records of these results.

Because NIOSH has only identified six documented releases from 1968-1974 (an 

average of 1 per year), the application of a daily release would be a 

significant/bounding overestimate of the number of RFP tritium releases.

3. Tritium was released to the workplace environment, and not in a glovebox.

4. The release involved elemental tritium (HT, T2), and not tritium oxide (HTO).

5. The tritium was released from a contaminated shipping container which was procured 

by Rocky Flats in 1970 and can be taken as representative of shipping containers in 

use prior to 1973.
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As stated in Item 2, the 1974 1.5-Ci tritium release is the only documented release 

from a shipping container in the Rocky Flats workplace. It is taken to be typical since 

there are no other such documented releases to use in forming the model. There is 

documented concern about such releases, as shown in the following quote from the 

ChemRisk report (ChemRisk 1994, pdf p. 38):

As early as 1962, Rocky Flats maintained instruments for detection of tritium gas in 

particular work areas of the plant because operations have sometimes resulted in the 

storage of tritium containers.

The instruments available to Rocky Flats at that time were only semi-quantitative for 

indicating the presence of tritium; NIOSH has captured no records of these results.

Because NIOSH has only identified six documented releases from 1968-1974 (an 

average of 1 per year), the application of a daily release would be a 

significant/bounding overestimate of the number of RFP tritium releases.

6. The incident occurred close enough in time to the 1973 tritium release that work 

practices and controls were likely more similar to those prior to 1973 than to those 

even a year or two later, as procedures and controls evolved with greater sensitivity to 

the potential for tritium contamination.

An assessment of the 1974 1.5-Ci tritium release from a contaminated shipping container on August 

30, 1974 was made based on the data in an RFP report (AEC ca. 1974). Specific urine sample 

collection dates were not included in the report, but data were matched to two claims in the NIOSHOffice of Compensation Analysis and Support Claims Tracking System (NOCTS), which reported a 

collection date of September 5, 1974. An assessment of this data was performed using an intake 

date of August 30, 1974, and the largest reported result collected after the incident (36,320 pCi/L). 

This resulted in a dose of about 0.15 mrem. Assuming one incident per workday at 0.15 mrem for 

250 workdays per year results in an annual dose of 37.5 mrem/yr for the pre-1973 period. This 

should be assigned to all unmonitored RFP radiation workers.

5.3.5.2 1973 Tritium Release Exposure Method

The report, Investigation of the Tritium Release Occurrence at the Rocky Flats Plant (AEC 1973), 

describes a 1973 incident that prompted the site to sample a number of workers for tritium exposure 

(examples include ORAUT 2012a, 2012b, 2012c). A shipment of scrap plutonium from the Lawrence 

Livermore National Laboratory (LLNL) was discovered to have been contaminated with tritium. This 

material was processed at the RFP from April 9 to 25, 1973, in Building 779A, Room 154. Because it 

was not immediately identified as being contaminated, monitoring of potentially exposed individuals 

did not begin until late September 1973.

Two hundred fifty people were sampled after the discovery; this included all employees who worked in 

areas in which the contaminated scrap was processed. The waste stream from the processing of this 

material was also contaminated, which resulted in the potential for intakes of tritium at later dates. 

Therefore, all employees who were involved in the processing of wastes from this scrap were also 

included in the urinalysis program. The collection of samples from a tritium-contaminated water 

bubbler on September 19 and September 25, 1973, were also identified as possible sources of 

intakes.

Due to the large sample load, raw urine samples were first analyzed in many of the cases. It was 

noted that the counting efficiency was only about 3% for these analyses, and that the corrections for 

spectral shift could lead to abnormally high readings. Nineteen employees were initially identified as 
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having elevated tritium levels in their urine. These samples were distilled and reanalyzed. This 

recheck found fourteen of these employees were below the 10,000-pCi/L action level at the site. The 

five most-exposed individuals were identified, and details of their potential exposures, including 

bioassay results, are included in the investigation report. One of these five individuals is in NOCTS. 

The results of the five workers who exceeded the 10,000 pCi/L action level were reviewed by NIOSH.

Exhibit 14 of the report contains a section on Personnel Exposure Data. The following is an excerpt 

(AEC 1973):

SAMPLING PROTOCOL

Dow began by sampling urine from all employees who were thought to have had the 

best chance of being exposed to tritium. As of October 15, 1973, about 

250 employees have been tested. Dow is continuing to trace leads to other possible 

exposure and will sample them as they are found. Dow intends to sample many 

employees who have had only a remote chance of coming in contact with tritium. Dow 

also tests the urine of any employee who requests this whether or not they are 

candidates for exposure.

ACTION LEVELS

An “action level” of 10,000 pCi/l was tentatively chosen for resampling. This level was 

chosen for several reasons such as:

1. An article by Fitzsimmons indicated that people wearing tritiated watches could 

excrete levels of 10,000 pCi/l.

2. A calculation of worst possible circumstances indicate that an employee would have 

to exceed levels of 23,000 pCi/l before any permissible yearly levels of whole body 

radiation would be exceeded.

3. The sample load was such that Dow could handle resampling only a limited number 

of employees on a frequent basis. It turned out that a relatively small number were 

over 10,000 pCi/l, but a large fraction were in the 5,000 and 10,000 pCi/l range.

4. Without predistilling the urine samples the counting efficiency drops to about 3% and 

the corrections made for spectral shift can lead to abnormally high reading.

5. With a large sample load, counting time devoted to each sample must be restricted 

so that 10,000 pCi/l might be considered lowest detection limit available under the 

present circumstances. All samples above 10,000 pCi/l are redone by counting the 

distillate of the original sample.

RFP identified five workers with tritium urinalysis results above the action level of 10,000 pCi/L. 

Results from these five workers are reviewed here. Fourteen other workers had results initially above 

10,000 pCi/L, but these fell below this level on recount (as noted above, the distillates of the original 

samples were counted, which offered better counting statistics during recount).

The document contains information, including tritium bioassay results and brief work histories, about 

the five workers with the largest tritium sample results. This information was used to assess the 

doses to the affected workers and is displayed in italics in the sections below. All five cases had initial 

samples that were not distilled, with one to five later samples that were distilled. In general, the 

undistilled and distilled sample results tended not to agree with the distilled samples, which yielded 
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 29 of 177

lower values. This is to be expected in light of item 4 in the action levels discussion above. The 

predistilled results were used in the development of this analysis because there were more results 

available and they yielded doses that are favorable to claimants. The following assumptions were 

used in this assessment:

• Equal weight of all samples (measurement error the same for all samples),

• Only predistilled samples for fits,

• Tritium in the form of tritiated water vapor (HTO),

• The Integrated Modules for Bioassay Analysis (IMBA) model for inorganic tritium, as described 

in Guidance on Use of IMBA Software for DOE Safety Applications (DOE 2006),

• Injection intake (for modeling with IMBA), and

• Intake dates based on worker information and examination of fit to urine sample results.

The five workers who had the largest tritium urinalysis results are assessed below. Text in italics 

indicates an excerpt from the incident report.

Case A

Case A worked in Room [location redacted] from [date range redacted].

He was involved in the hydrating [sic: likely hydriding] and processing of the parts in 

question from [date range redacted], along with Cases [case identifiers redacted]. He 

was not involved in any of the following special projects:

a. [date, special project name redacted]

b. [date, special project name redacted]

c. [date, special project name redacted]

He was involved in taking samples from a tritium-contaminated [device redacted] on 

[dates redacted]. On [date redacted], this was done without a [item redacted].

From this history, it would appear the most likely exposure occurred on [dates 

redacted]. If an exposure had occurred between [date range redacted], it is likely that 

both Cases [case identifiers redacted] would have been exposed to the same source, 

and subsequently, excreted the same quantities of tritium.

The RFP document also states:

In Case A, a history of his work assignment and his urine results for the first two weeks 

indicate that he sustained a recent exposure. At the present time he is excreting tritium 

with an elimination half life of less than 10 days. According to Sanders and Snyder, 

this is the pattern of elimination from an exposure up to 90-days post exposure.

The statement that Case A’s intake appears to be recent agrees with current models for HTO intakes. 

If an intake on [date redacted] is assumed, a very poor fit to the data is achieved.
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Based on the worker’s history and the bioassay result pattern, an acute intake was assumed on 

September 19, 1973. Using the results of samples from September 25 to October 4, and applying a 

uniform error to each of the samples, the intake is 38.7 μCi. The corresponding dose is 2.6 mrem. 

These samples are presumed to be predistilled because later samples from October 5 to 12 are 

labeled as “distilled.” This yields a very good fit to the predistilled results.

Case B

He has worked in [location redacted] since [date redacted]. He was in the room when 

[action redacted].

Assuming a chronic intake from July 1 through September 25 (date of first urine sample) yields an 

intake rate of 0.33 μCi/d (for a total intake of 28.1 μCi) and provides a reasonable fit to the results. 

The dose is 1.90 mrem.

Assumption of an acute intake on September 19 (date of the first bubbler sample) yields an intake of 

7.28 μCi. This fit is almost identical to the first scenario.

A single acute intake on the first day in the area (July 1) yields an intake of 720 μCi and a dose of 

49 mrem.

The single acute intake on July 1 does not provide a good fit to the later predistilled results. The first 

two scenarios (chronic intake from July 1 through September 25, and acute intake on September 19) 

provide similar fits that reasonably follow the pattern of the predistilled samples. The chronic intake 

yields a larger intake so it is used for the best estimate.

Case C

He worked in [location redacted] since [date redacted]. He was not in the room when 

[action redacted].

Because the worker did not start in the area until August 27, an acute intake was assumed on this 

date. Using only the predistilled sample results, the intake is 21.3 μCi with a dose of 1.4 mrem.

If a chronic intake is assumed to have started on his first day of potential exposure (August 27) and 

continued until the date of his first sample (September 25), the resulting intake is 0.24 μCi/d for a total 

intake of 7.08 μCi.

The two fits are very similar, so the acute intake is selected as the best fit because it results in a dose 

that is more favorable to the claimant.

Case D

He worked in [location redacted], between [date range redacted]. He has not been 

exposed to tritium since [date redacted].

Case D submitted samples on only three days, although there are two results on two of those days. 

In one instance, one of the samples was distilled; on the other day, there is a note stating “repeated 

with sample channel ratio.” On the latter day, the results differ by a factor of almost 2; the larger of 

these results is assumed to be the predistilled analysis and is used for the intake assessment. An 

assumed chronic intake from April 10 through 25 (last date of incident) yields an intake of 71.2 μCi/d 

for a total intake of 1,070 μCi. The resulting dose is 72 mrem.

A chronic intake from April 10 to June 15 yields an intake rate of 8.84 μCi/d for a total intake of 

581 μCi (39 mrem).
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Because there are few samples and the results follow no specific pattern, there is little difference 

between the fits. Therefore, the acute intake is assigned because it yields the larger dose.

Case H

He came in contact with the possible source of tritium on [date redacted].

No other information is included in the report. The conclusion in the report is:

It is expected that, as a result of a review of his work history and urinalysis data, a dose 

assignment of less than 3 rem will be made.

However, no follow-up information is available. Because the only available information indicates that 

an intake would have occurred on April 6, an acute intake was modeled. The resulting intake is 

1,240 μCi with a dose of 84 mrem.

The best estimates for the five reviewed cases are summarized in the Table 5-12 below.

Table 5-12. Summary of RFP tritium dose estimates.

Case

Intake

(μCi)

Dose

(mrem)

A 38.7 2.6

B 28.1 1.9

C 21.3 1.4

D 1070 72

H 1240 84

The tritium contamination was associated with plutonium scrap material. Therefore, the largest 

assessed dose of 84 mrem should be assigned to all individuals who were monitored for plutonium in 

1973.

5.3.5.3 Post-1973 Unmonitored Tritium Exposure

For the assessment of tritium exposures at RFP after 1973, an analysis of NOCTS data from 1974 

and 1975 was performed. There are 38 individuals with tritium data in 1974 and 37 in 1975. ORAUTOTIB-0075, Use of Claimant Datasets for Coworker Modeling, provides justification and guidance 

(ORAUT 2009).

When assessing tritium intakes for most sites, it is assumed that intake potential exists only during 

tritium bioassay monitoring because monitoring is cheap, easy, and requires only spot samples, and 

therefore presents less of a burden than other forms of bioassay on both the employer and the 

employee. Because tritium was not of primary concern at RFP and was present only as a potential 

contaminant on equipment, a particular individual was not placed on a routine sampling program. 

Instead, a program was established in which one-tenth of the collected urine samples for plutonium 

analysis were also analyzed for tritium content (Bowman 1974) as well as the collection of samples 

when there was a particular concern. Samples available in NOCTS for these two years indicate that 

analyses were performed throughout the year, with most individuals being sampled only once.

For the coworker study, it was assumed that each worker had the potential to be exposed at a 

constant level throughout the year in which the urine sample was collected. The 95th percentile was 

used because one-tenth of the population was sampled. The coworker study for 1974 to 1975 yielded 

doses of 0 mrem for everyone.

For the years after 1975, there are 11 or fewer individuals in NOCTS with tritium data; this is 

insufficient for performing a coworker study. Results for these years are consistent with those from 
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the previous years and show a general decreasing trend. The intake rate from the 1974 to 1975 

coworker study (i.e., 0 mrem) applies to these years. Therefore, no additional unmonitored tritium 

dose should be assigned after 1973.

5.3.5.4 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Starting in 1973, workers were monitored for possible tritium exposures only for special projects or 

situations. The methods have not been reviewed but probably involved liquid scintillation 

measurements [34]. The urine results are reported as picocurie per liter of urine, and actual results 

were reported, generally with the standard deviation. It has not been determined whether the 

reported uncertainty in the 1970s to early 1980s is 1 or 2 times the standard deviation. The sensitivity 

of the method was 2 to 3 orders of magnitude better than the significant level of about 1 μCi/L. 

Although the actual MDA has not been quantified for the methods in the 1970s and 1980s, it is 

probably in the range of several hundred to several thousand picocuries per liter [35]. The MDA for 

tritium should be assumed to be 600 pCi/L for all tritium bioassay (RFETS 1998c, p. 176).

5.4 IN VIVO

In vivo lung counts have been performed at RFP since 1964 to determine the activity of plutonium in 

the lungs of workers who were exposed, or had the potential to be exposed, to airborne plutonium. 

The method of in vivo lung counts was to place one or more detectors over the chest of the subject 

and count the photons that are emitted from the plutonium mixture, if any, in the chest. Plutonium 

was not detected directly because of the low abundance of gamma photons and the severe 

attenuation of the more abundant low-energy X-rays (L X-rays). Rather, the 59.5-keV gamma photon 

from 241Am was used to detect 241Am, which is present to some extent in all WG plutonium at RFP. 

The activity of plutonium was then calculated from the detected 241Am by measuring, calculating, or 

assuming the fraction of the 241Am in the plutonium mixture on the date of the lung count (see 

Section B.11 in Attachment B). At RFP, the fraction of the 241Am in the plutonium mixture has 

historically been characterized in terms of parts per million by weight. Direct in vivo measurement of 

plutonium in the lungs, although investigated, was never implemented at RFP (Falk et al. 1979).

The RFP lung counter also measured 234Th, using the 63-keV gamma (doublet) photon, to determine 

the activity of 238U in the lungs of workers exposed to DU. This measurement was made possible by 

the improved resolution of the germanium detectors that allowed baseline separation of the 59.5-keV 

gamma of 241Am from the 63-keV gamma doublet of 234Th. The activity of 238U was considered to be 

equal to that of the measured 234Th under the assumption of equilibrium (Berger 1988a).

Attachment B, Minimum Detectable Activity for In Vivo Lung Counts at RFP, contains more detail. 

Section 5.6 discusses the data, and Attachment C contains examples of the report forms.

5.4.1 Americium and Plutonium

5.4.1.1 Methods, Units, Isotopes, and Interferences

Before April 1997, lung count data were not converted to a quantified amount or activity unless there 

was confirmation that the count was from an actual deposition in the lungs. For unquantified results, 

the data are generally in units of counts per minute and accompanied by a decision that is noted as 

normal, background, or some abbreviation of background. For quantified results through about 1968, 

the unit was micrograms of plutonium. In addition, the result was converted to a fraction of the 

maximum permissible lung burden (MPLB) using a plutonium-specific activity of 0.07 μCi/μg and the 

MPLB of 0.016 μCi (16 nCi) for the alpha-emitting isotopes of plutonium. Starting in about 1973, the 

activities of both plutonium (including all the alpha-emitting isotopes of WG plutonium) and americium 

(

241Am) were recorded in nanocuries [36]. In addition, the activity of 241Am was stated as a fraction of 
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the MPLB, which was 14.7 nCi (Falk 1993). After 1989, the results were no longer stated as a fraction 

of the MPLB.

There are two sources of interferences to consider. The first is the 63-keV gamma doublet of 234Th 

from DU operations being mistaken for 241Am in lung counts with the NaI or phoswich detector 

systems. This interference is most troublesome to dose reconstruction for workers with residual lung 

depositions of plutonium and americium who subsequently worked in DU operations [37]. The second 

interference is the contribution of count from 241Am not in the lungs (e.g., contributions from 

contamination on the skin, from material being cleared from the upper respiratory system, or from 

ingested material). A positive detection of 241Am did not necessarily indicate an intake (especially one 

that resulted in a deposition to the alveolar-interstitial region of the lungs) of the plutonium-americium 

mixture, especially for a lung count in response to an incident [38].

5.4.1.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Reporting levels are not easily defined because quantification was preceded by verification counts 

and professional judgments. In addition, before 1974, the practice was not to quantify a positive 

detection of 241Am unless the deposition could be associated with a known incident with a known ppm 

241Am. Affected workers were classified as positive unknowns or some variation. Starting in 1974, 

the practice was changed to quantify the plutonium depositions for positive unknowns by assuming a 

default value of 1,000 ppm 241Am on the date of the most probable intake or on the date of the first 

positive lung count. The ppm 241Am was then calculated for the date of the lung count to account for 

the ingrowth of 241Am from the nuclear transformation of 241Pu and the radioactive decay of the initial 

241Am [39].

In general, this quantification was not applied retroactively to earlier positive lung counts. Once a lung 

deposition of plutonium had been quantified for a worker, the deposition continued to be quantified for 

all subsequent lung counts (except screening counts for new intakes), regardless of the result of the 

subsequent lung count (including negative values), until each of the last three results was less than 

the decision level for the count and the average of the last three results was within 1 standard 

deviation of 0.00 nCi plutonium [40].

The decision levels varied. From 1965 to 1968, the decision level was two times the uncertainty of 

the matched subject’s net count, although the application of this decision level was inconsistent in this 

period. Starting in 1969, for NaI and phoswich detector systems, the decision level was 3 times the 

standard deviation of the net count rate for a set of lung counts for unexposed known cold subjects 

based on the index method (see Attachment B). Results between 2 and 3 sigma were noted but not 

always investigated. For the germanium detector systems, starting in 1976, the decision level (also 

called the “cutoff”) was equal to 1.645 times the standard deviation of the net count rate [41]. The 

decision level for 1995 and later was calculated by ABACOS-Plus for a probability of a Type I (false 

positive) error of 5% (RFETS 2000b, p. 90). The decision level was used as a reporting level from 

1995 to early 1997.

Table 5-13 lists the MDAs for 241Am, which were calculated for the evolution of lung-counting systems 

at RFP as described in Attachment B.
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Table 5-13. Summary of MDAs for 241Am.

Perioda Detector system Index

MDA (nCi) for 241Amb

Minimum system Standard system

Half time Full time Half time Full time

1964–1968 NaI(Tl) 4 × 4 0.90 1.7 1.5 1.3 1.2

1.35 2.8 2.5 2.1 1.9

1.80 4.6 4.1 3.5 3.2

1969→ NaI(Tl) 4 × 4 0.90 – – 0.8 0.76

1.35 – – 1.3 1.3

1.80 – – 2.2 2

1973→ Phoswich 0.90 – – 1.2 1.2

1.35 – – 2.0 2.

1.80 – – 3.3 3.2

1976–1978 Ortec Arrays

(High-purity Ge)

0.90 0.26 0.18 0.2 0.14

1.35 0.48 0.32 0.37 0.25

1.80 0.86 0.59 0.66 0.45

1979→ Ortec Arrays

(High-purity Ge)

0.90 0.2 0.14 0.16 0.11

1.35 0.37 0.25 0.28 0.19

1.80 0.66 0.45 0.51 0.35

1978→ PGT I Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.4 0.27 0.31 0.21

1.80 0.71 0.49 0.55 0.38

1979→ PGT I Arrays

(High-purity Ge)

0.90 0.17 0.12 0.13 0.09

1.35 0.31 0.21 0.24 0.16

1.80 0.55 0.38 0.42 0.29

1979→ PGT II Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.4 0.28 0.31 0.21

1.80 0.74 0.5 0.57 0.39

1985→ PGT Organ Pipe Ge Detectors 0.90 – – 0.15 0.11

1.35 – – 0.26 0.18

1.80 – – 0.46 0.32

1991→ EG&G Organ Pipe Ge Detectors 0.90 – – 0.14 0.1

1.35 – – 0.26 0.18

1.80 – – 0.48 0.33

1995→ Ortec 2 Organ Pipe Ge Detectors 0.90 – – – 0.14

1.35 – – – 0.3

1.80 – – – 0.6

a. In overlapping years the more favorable MDAs should be assumed.

b. – = not applicable.

These values of MDAs are for three indices that represent the median and the approximate 5th- and 

95th-percentile body statures of RFP male workers. To obtain the worker-specific MDA, dose 

reconstructors can calculate the value using the information in Attachment B or interpolate (or 

extrapolate) from the values in Table 5-13 [42]. The worker-specific index is generally stated on lung 

count report forms from 1969 to 1994 and can be derived from the weight and height data on report 

forms from 1995 and later. (The MDA values are reported on report forms from 1995 and later, but 

the values are not worker-specific. Dose reconstructors should disregard these MDA values.) The 

default MDA would be for an index of 1.35 if height and weight (or index) data for the worker are not 

available [43]. The DR should assume the MDA is twice the decision level for 1995 and later lung 

count reports that include the non worker-specific MDA.

The MDA for plutonium should be calculated by multiplying the worker-specific value of the MDA for 

241Am by the MDA conversion factor (Equation B-17 in Attachment B), which is based on the value of 

the ppm 241Am on the date of the lung count. The value of the ppm 241Am on the date of the lung 

count, accounting for ingrowth of 241Am from the nuclear transformation of 241Pu and the radioactive 

decay of the initial 241Am, is given by Equation B-18 in Attachment B. Dose reconstructors need to 

establish the date of the intake and the initial ppm 241Am. If that information is not apparent in the 
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available records, an approach that is favorable to claimants is to assume the initial ppm 241Am to be 

100 [44].

The assumption of the intake date is not straightforward and should balance maximizing the plutonium 

lung deposition (intake date is close to the date of the lung count) and maximizing the accrued lung 

dose (intake date is far from the date of the lung count). In addition, the choice of intake date for the 

lung count data should be coordinated with that for the associated urine data [45].

Dose reconstructors must choose the value of the initial mass fraction of 241Pu. At the RFP lung 

counter, 0.005 was historically used as the initial mass fraction of 241Pu and is a realistic choice for 

intakes that occurred in the 1950s to June 1976. The fraction 0.0036, based on the isotopic 

composition for RFP stream plutonium in the mid-1970s, should be used for intakes that occurred 

from July 1976 to 1989. For intakes after 1989, the initial fraction of 241Pu should be reduced to 

account for the aging (radioactive decay) of the 241Pu [46].

The uncertainties of the results were reported for the net counts per minute starting with the 

germanium detector systems in 1976. The uncertainty was reported at 1 standard deviation and 

included only the contribution from counting statistics. Starting in approximately 1981, the counting 

statistics uncertainty was also applied to the assessed activity and to the value of the fraction of the 

MPLB [47]. With the advent of ABACOS-Plus in 1995, the percent error at 1 standard deviation was 

reported for all identified nuclides. Beginning on October 11, 1999, a 30% systematic uncertainty, 

which included contributions of uncertainties in the chest wall thickness (CWT), the location of the 

activity in the lungs, the uncertainty in the ppm 241Am, and the influence of activity deposited in other 

organs, was included in the total propagated uncertainty (RFETS 2000b, p. 89).

The major uncertainty for the calculation of the plutonium lung deposition is the ppm 241Am in the 

plutonium in the lungs at the time of the lung count. Factors in the uncertainty are the intake date, the 

value of the initial ppm 241Am, the initial fraction of 241Pu, and the degree of association of the 

americium with the plutonium while in the lungs. An underlying assumption is that the americium 

remains associated with the plutonium particles in the lungs until the particles are dissolved or 

removed from the lungs. The degree of validity of this assumption has not been determined [48].

5.4.2 Thorium and Depleted Uranium

5.4.2.1 Methods, Units, Isotopes, and Interferences

The method to detect DU was to detect the 63-keV gamma (doublet) photon of 234Th and to calculate 

the activity of 238U assuming equilibrium. This method was implemented manually for special cases in 

approximately 1978. Starting in 1983, the count data for the 63-keV doublet photon were routinely 

processed and reported. However, the activity of the 238U was calculated only for special cases and 

not routinely. A supplemental method, implemented in about 1989, detected the 93-keV gamma 

doublet photon of 234Th, and the count data were routinely processed and reported. This

supplemental method was used mainly to reduce false positive results for the detection of 234Th 

because detection of both doublet photons was required before detection of 234Th was considered.

Starting in 1995, the activity of 238U was calculated and reported if the 63-keV peak (or sometimes the 

93-keV peak) was detected by the ABACOS-Plus peak-search software. If the peak was not 

detected, the activity of 238U was reported as less than the decision level (the activity of the decision 

level was reported). Starting in early 1997, the activity of 238U was reported, including negative 

results, even if a peak was not detected. In a similar manner, the activity of 235U was reported. 

Starting in about 1999, the activity of 238U was based solely on the 63-keV peak.
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The main part of the data for the 63-keV doublet photon is in units of net counts per minute. To 

convert to activity (nanocuries) of 238U, the counts per minute is divided by the calibration factor for 

241Am (see Attachment B) and normalized to the ratio of photon abundances [abundance of 59.5-keV 

gamma, 241Am, is 0.359; abundance of 63-keV doublet gamma, 234Th, is 0.0381 (Lederer and Shirley 

1978); the ratio (59.5-keV gamma/63-keV doublet gamma) is 9.4]. That is, nanocuries of 238U equals 

[(234Th 63-keV net cpm) divided by (241Am calibration factor)] multiplied by 9.4. To calculate the 

activity for DU, the 238U activity is divided by 0.89 (see Section 5.2.3.2.1).

The interference is 238U in natural uranium. Unless there is a reported activity for 234U that is 

approximately equal to that reported for 238U, dose reconstructors should use the assumption 

(favorable to claimants) that the 238U activity is all from occupational exposure to DU [49].

5.4.2.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Reporting levels were not generally used for DU until 1995 with the implementation of ABACOS-Plus 

(see Section 5.4.2.1). Before 1995, the 238U activity was generally quantified only after verification of 

an intake.

The MDA for 238U has not been determined rigorously. However, the 238U worker-specific MDA can 

reasonably be expected to be a multiple of the 241Am worker-specific MDA because the detected 

photons (63 keV and 59.5 keV) are very close in energy. As described in Section 5.4.2.1 for using the 

calibration factor for 241Am to determine the 238U activity, the 238U worker-specific MDA can be 

obtained by multiplying the 241Am worker-specific MDA by 9.4. That result is divided by 0.89 to obtain 

the worker-specific MDA for DU [50]. (As noted in Section 5.4.1.2 for americium and plutonium, MDA 

values are reported on forms for 1995 and later, but are not worker-specific. Dose reconstructors

should disregard these MDA values.) The DR should assume the MDA is twice the decision level for 

1995 and later lung count reports that include the non worker-specific MDA.

The major uncertainty is the assumption of equilibrium of the 234Th with the 238U before 1990, when 

DU was still being processed. Part of the process was to remove decay chain radionuclides, 

especially thorium, by heating the uranium ingot to drive the smaller atoms of thorium to the surface or 

top of the ingot, which was then cut off. The result was DU metal with a deficiency of 234Th for several 

weeks plus scrap DU with an excess of 234Th (super-equilibrium). The assumption of equilibrium 

when super-equilibrium existed is favorable to claimants [51]. The effect of a deficiency of 234Th (not 

favorable to claimants) is mitigated by the rapid ingrowth of the 234Th into the DU. Fifty-percent 

equilibrium occurs after 24 days after a thorium strike, and 90% occurs after 80 days.

The standard deviation of the net count rate is reported through 1995 but includes only the 

contribution of counting statistics. To estimate the uncertainty of a 238U or DU activity from the net 

count rate, dose reconstructors can divide the worker-specific MDA by 3.3 [52].

5.5 OTHER BIOASSAY DATA

5.5.1 Wound Count Data

Wounds are defined as any break in the skin (e.g., cuts, punctures, abrasions, acid burns). Any 

wound that occurred in a work area involving plutonium was monitored for plutonium contamination, 

especially after the advent of the wound counter in 1957. Counting a blood sample or directly 

counting the wound site with an alpha detector were also methods RFP used to monitor wounds to 

detect possible plutonium contamination. In RFP terminology in the 1950s and 1960s, wound counts 

were called “gamma specs,” and the wound counter was called a “gamma spectrometer.” Wounds in 

uranium work areas were monitored selectively. The record could contain an incident report, a wound 
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count data sheet, a medical decontamination report, and a medical treatment report, depending on 

the era and circumstances.

The process was to attempt to decontaminate the wound in the building of the occurrence by washing 

and encouraging bleeding to flush any plutonium out of the wound. Then the worker was sent or 

escorted to the medical facility for a wound count and additional decontamination if the wound count 

was positive (Berger 1988b). The sequence of additional decontamination was washing with soap 

and water, washing with commercial bleach, scrubbing with commercial bleach, and excision.

Wound count information is largely irrelevant to dose reconstruction [53]. The relevant items are the 

urinalysis data, the identification of the mode and date of intake, and whether there was residual 

plutonium at the wound site. Guidance on assessing wound intakes is provided in Technical 

Information Bulletin: Guidance on Wound Modeling for Internal Dose Reconstruction (ORAUT 2005)

5.5.2 Nasal Smears and Fecal Samples

Nasal smear (later called swab) and fecal sample data were occasionally performed throughout RFP 

operations as supplemental data for workers with actual or suspected significant inhalation intakes. 

Through the 1980s, they were used subjectively to verify that an intake did occur and to estimate the 

possible magnitude of the intake. The data were also used to determine or confirm the ppm 241Am in 

the inhaled plutonium mixture. Some obstacles to using nasal smear or fecal data to quantify an 

intake are unknown particle size distribution, unknown fraction of the plutonium captured by the nasal 

smear or fecal sample, inconsistent and largely undocumented sampling technique for nasal smears 

(which sometimes were called “nose blows”), and unknown counting efficiency (e.g., sample geometry 

and alpha absorption, especially in the 1950s and 1960s). Through 1989, the requested fecal sample 

was the second voiding after the incident. In some cases, the second, third, and fourth voidings were 

requested.

Starting in the 1990s, the nasal or mouth smears were used as a workplace indicator to identify 

potential intakes, and fecal sampling was used to confirm and evaluate suspected intakes (RFETS 

1998d, p. 62).

The reported MDAs (RFETS 1998d, pp. 67–68) are:

• 20 dpm/sample, for (gross alpha, liquid scintillation) routine nasal samples;

• 0.2 dpm/sample, for fecal samples with a 21-day reporting time (plutonium alpha isotopic);

• 1.3 dpm/sample, for fecal samples with a 14-day reporting time (plutonium alpha isotopic);

• 2.6 dpm/sample, for fecal samples with a 7-day reporting time (plutonium alpha isotopic); and

• 100 dpm/sample, for fecal samples with a 2-day reporting time (nonisotopic, rapid analysis).

The reporting times are the times for the laboratory to analyze the sample and report the results. The 

shorter reporting times indicate an expedited analysis, with the trade-off of a less sensitive analysis (a 

higher MDA).

These MDA values apply to samples starting approximately in 1993 and are specifications for the 

laboratory. (Note: The laboratory MDA does not depend on the time after intake that the sample was 

excreted.) Most reports of fecal sample results do not give the sample-specific MDA but might give 

the decision level (Lc), which is approximately one-half of the sample-specific MDA. MDA values for 

earlier years are not available.
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5.6 RECORDS AND REPORTS

This section discusses the interpretation of the data and information on records and reports of 

bioassay data [54]. Attachment C, Examples of Records and Reports Used at RFP, contains the 

figures described below.

5.6.1 Urinalysis Records and Reports

Figures C-1 to C-3 are examples of the Urinalysis Record Card and the HSDS – Urinalysis Detail 

report. The Urinalysis Record Card was the recording medium for the urinalysis data from 1952 to 

1969 and is the primary record for urine data in this period. The urine data were manually entered on 

this card through 1969. These data were also entered into a database starting in about 1961. In 

about 1970, the HSDS was implemented to record, process, and report urinalysis data and the 

derived fraction of the maximum permissible systemic burden.

5.6.2 Interpretation of the Urinalysis Record Card

Urine results are presented in columns under the month for a given year (in the row). The top number 

is the day of the month (assumed to be the excretion day). The middle number is the sample result, 

either a number or BK (see Section 5.3.1.2). The bottom number is the technique code and refers to 

the codes in the header (see Attachment A).

The unit of the result is given in the header. Sometimes the unit is written with the result (e.g., μg in 

Figure C-1, analysis Code A, 1955). Be careful not to interpret μg as the number 49.

The corresponding data on the HSDS – Urinalysis Detail report should be the same as that on the 

Urinalysis Record Card. If not, the data on the Urinalysis Record Card should be taken as the correct 

data, with the exception noted in Section 5.3.1.2 (i.e., some plutonium results reported as BK on the 

card were rereported with the actual result) [55]. On some cards, dose reconstructors might observe 

the initially reported result was crossed out and replaced by a lower value. The technical basis for 

that change has not been determined. In addition, that change generally was not applied to the data 

in the HSDS. It is reasonable and favorable to claimants to disregard the modified result [56].

The analyte code for DU was sometimes transcribed incorrectly from the card to the urinalysis detail 

report as U (see Figure C-1) rather than D (see Figures C-2 and C-3) with the unit of dpm/24-hr

sample rather than μg/24-hr sample.

Figures C-4 and C-5 are two versions of urinalysis reports from the HSDS. Both versions report the 

data in the same way but with differences in the headers. Figure C-5 (the newer version) adds a 

column (the uncertainty of the result).

5.6.3 Interpretation of the Health Sciences Data System – Urinalysis Detail Report

The Activity Date is taken to be the date that the sample was excreted. However, the recorded date 

frequently was the date that the sample was received at the laboratory, especially for routine samples. 

(This applies also to the dates on the Urinalysis Record Card.)

ANAL is the code for the analyte:

P = plutonium,

A = americium,

U = EU (pre-1970, approximately),

U = DU (1970–1989, approximately),
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D = DU (1952–1969, approximately), and

G = gross alpha.

NO CAL is a code used to flag the logic of the software.

0 = use normally in the calculation;

1 = do not use in the calculation; and

2 = date of a new intake.

Code 1 was used primarily for two situations to exclude a sample result from the systemic burden 

calculation: If the excretion of the analyte was enhanced by a chelation treatment or if the analysis of 

the sample did not meet quality standards (an invalid analysis or result). Sample results within 

90 days of a chelation treatment were generally (or should have been) coded as 1 [57]. The use of 

Code 2 to flag the date of a new significant intake occurred inconsistently. In reports from the 1980s, 

an asterisk was used instead of a Code 2 to flag the date of a new intake. Dose reconstructors

should disregard the Code 2 or asterisked entries.

• ELAPSED DAYS is the number of days since the hire date. This data field is not likely to be of 

use.

• The EXPOSURE VALUE or DPM/24HR is the result of the urinalysis for the analyte. In 

general, the unit was dpm/24-hr sample, except for DU, from 1952 to April 1964.

• The column in parentheses is the uncertainty, starting in 1980. Any value or symbol in the 

parentheses before 1980 is only a placeholder and should be disregarded [58].

• The BODY BURDEN % or SYSTEM BURDEN is the fraction of the maximum permissible

systemic burden that was calculated from Code 0 results for plutonium and for americium. 

This data field is not likely to be of use.

5.6.4 Interpretation of Other Urinalysis Reports

Figures C-6 and C-7 are examples of urinalysis reports from the onsite bioassay laboratory from 1990 

to the mid-1990s. Figure C-6 is for a special urine sample for plutonium analysis, and Figure C-7 is 

for a routine urine sample for plutonium analysis. Both forms have the same format. The first three 

columns are self-explanatory; the remaining columns are:

• Dec Level is the decision level (Lc) in units of dpm/sample.

• Aspec is code for the alpha spectrometry quality. The Aspec codes are defined on the lower 

left portion of the report. Aspec code 0 is analogous to the previous Code 0 for urine data in 

the HSDS. Codes 1, 3, and 4 indicate a failed analysis and disqualify the result [59].

• DQO, for “data quality objective,” is the code for status of the data quality objectives for the 

results of the batch blank and control samples. The DQO codes are defined on the lower 

center portion of the report. DQOs, in theory, were assessed for the blank, accuracy, and 

precision. In practice, the DQO was usually assessed only for the blank. Therefore, the ANN 

notation means that the blank was acceptable, the accuracy was not assessed, and the 

precision was not assessed. An F would indicate that the batch failed a DQO. If the batch 

failed, every sample in the batch was conditionally failed pending further evaluation [60].

• Batch Val is the overall validation of the result. V means valid, and I means invalid. Do not 

use a result that has an I validation code [61].
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• Analyte is self-explanatory.

• Recovery is the fraction of the tracer recovered by the analysis.

• DPM is the result of the sample in dpm/sample. Dose reconstructors should assume a 

24-hour urine sample unless there is information that indicates otherwise [62].

• Error is the uncertainty at 1 standard deviation.

Figure C-8 is an example of the urinalysis data report by Quanterra, a commercial offsite laboratory, 

starting in 1993. The form header information, except for the collection date and the matrix, is not 

useful. The collection date, if not the sample excretion date, should be replaced by the sample date 

written on the form [63]. The result header is largely self-explanatory.

• The primary information is the RESULT and its TOTAL ERROR (at 1 standard deviation) in 

dpm/sample (REPORT UNIT).

• The decision level (Lc) and the sample-specific MDA are also stated.

• The YIELD is the percent recovery of the tracer.

• The RST/MDA is the ratio of the result and the sample-specific MDA.

• The RST/CNTERR is the ratio of the result and the counting error.

• The ANALYSIS DATE is the date the sample was analyzed, not the excretion date.

• The ALIQUOT SIZE is the volume of the sample in milliliters (ALQ UNIT).

• The DETECTOR ID is self-explanatory.

• The METHOD NUMBER references the document number of Quanterra’s analytical procedure 

used to process the sample.

Figures C-9 and C-10 are examples of the analytical report of the onsite bioassay in the mid-1990s. 

Most of the information is self-explanatory. Some points:

• The date sampled is the excretion date.

• The data can only be used if the Alpha Spec Condition Code is 0 and if the Data Validation 

Code is V.

• The 234U activity is approximately equal to 238U activity in Figure C-9, and both results are 

greater than the decision level. As stated in Section 5.3.3.1.1, this is the classic pattern 

indicating natural uranium, not an occupational intake of DU.

Figure C-11 is an updated version of the urinalysis data report of Quanterra. The significant 

improvement is the validation of each result (QUAL is V). Use only results with a QUAL of V.

Figure C-12 is the urinalysis data report for General Engineering Laboratories. The header 

information is largely self-explanatory.
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• The Date Collected is the sample excretion date. The 24-hour clock time (0600) is also noted; 

0600 was used as a default end time of the 24-hour excretion period if the actual end time was 

not documented [64].

• The VF is the volume fraction, the fraction of the sample that was analyzed. A VF of 1 

indicates that the entire sample was analyzed.

• Use only data that have a Data Validation Code of V.

Figure C-13 is an example of the data card that was used in the 1970s and 1980s to record data 

manually for tritium urine samples and for other samples such as fecal samples and nasal smears. 

The unit of the tritium results is pCi/L. The unit of the fecal sample and nasal smear results is 

dpm/sample [65].

There might be other versions of in vitro bioassay reports. In all cases, the important data are the 

excretion date, the analyte, the result in the proper units, and whether the result was valid.

5.6.5 Lung Count Records and Reports

Figure C-14 is an example of an early lung count report. The aftermath of the October 15, 1965, 

plutonium fire in Buildings 776 and 777 was the first extensive use of the lung counter to detect 

americium and plutonium depositions for RFP workers.

• The in vivo lung-counting system was called the Body Counter. In RFP terminology, the lung 

count was called a body count through 1989. Most claimants will probably use the term “body 

count” instead of “lung count.” Dose reconstructors should not mistake the RFP “body count” 

for a whole-body count, which was widely used at other facilities to detect intakes of fission 

products.

• The Time field was used either for the time of the day at the start of the count or for the length 

of the count. In this case, the length of the count was noted (40 MLT means 40 minutes live 

time) [66].

• The “Minus Bkg + match” notation indicates that the result is the net count rate after the room 

background count rate and the net count rate of a matched person was subtracted.

• The “1.4 LB” notation is the calculated plutonium deposition in terms of the multiple of the 

MPLB of plutonium (1 MPLB = 16 nCi plutonium alpha emitters) [67].

• The Body Location is the position of the detector. In this case, the detectors were over the 

right and left portions of the chest. In many early counts, one of the detectors was over the 

liver or gut or below the sternum rather than over one side of the chest. Those data have little 

dosimetric use [68].

Figure C-15 is the August 1967 revision to the Health Physics Body Counter Information form. The 

change was to present the results after subtraction of the room background [Net (1) c/m] and after 

subtraction of matched subject net cpm [Net (2) c/m]. In addition, the plutonium deposition was stated

in terms of micrograms of plutonium.

Figure C-16 is the August 1968 revision to the Health Physics Body Counter Information form.

• The Net cpm is the subject’s total count rate minus the room background count rate.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 42 of 177

• The Predicted cpm replaced the net count rate of the matched subject.

• The Result is the final net cpm.

• In this example, there is no measurement for the right chest. Dose reconstructors should 

estimate the contribution for the right chest before using data from this count, because the 

lung data set generally includes contributions from both right and left lungs.

Figure C-17 is an example of a lung count with no tabulated result. This is an example of a positive 

unknown case (see Section 5.4.1.2). In addition, note the tabulation of the index, which was used 

later to estimate the chest thickness. Sufficient information is presented here and in Attachment B to 

allow dose reconstructors to calculate the plutonium and americium activities for this lung count, for 

any assumed or actual intake date.

Figure C-18 is the December 1973 revision to the previous form, with expanded information.

• The ROOM is the designation of the counting chamber, A, B, or C, used for this count.

• The RATIO field is the ratio of the 241Am photopeak region of interest (ROI) and a background 

ROI around 100 keV. The ratio was used as a supplemental subjective tool to improve 

detection of americium. A ratio of 1.20 or greater indicated probable detection of americium 

[69].

• The ppm 241Am was used to record either the ppm 241Am for a new incident or, as in this case, 

the calculated value of the ppm 241Am (including of 241Am) for a previous actual or assumed 

intake.

The form included fields to record the activity and fraction of the MPLB for both plutonium and 

americium. (This lung count, now quantified, is for the same positive unknown case as Figure C-17).

Figure C-19 is an example of the previous form for a count that was judged to be background. Data 

fields were added to capture data for measurements of the L X-ray (17-keV) ROI, especially for the 

phoswich detector system. Although that information was captured occasionally, the data were not 

used because of the instability of the predicted background cpm [70].

The previous lung count reports were for counts using the NaI detector system. Figure C-20 is an 

example of the lung count data for a germanium detector system. The data for the five to eight 

detectors of the germanium systems were multiplexed into a composite total count tabulated in the 

row for TOTAL CHEST. The standard deviation of the resultant counts per minute is based only on 

counting statistics. For workers with confirmed lung depositions, the calibration factors for plutonium 

and americium were generally written on the form, as in this case.

Figure C-21 is an example of the first computer report for the lung count results. The data are labeled 

appropriately. This report is for a worker with a confirmed deposition. The report for workers without 

a confirmed deposition does not report the calibration factors, the ppm Am, or the lung burden. 

Rather, it reports the cutoff, which is the decision level, and Normal if the DIFFERENCE is less than 

the cutoff [71].

Figure C-22 is an example of a computer report for the phoswich detector system, which was used as 

a backup screening system in the 1980s. Note the outcome statement, RESULTS ARE NORMAL. If 

the results were not normal, the subject would have been recounted with a germanium detector 

system [72]. Because the phoswich system could not resolve the 60- and 63-keV photopeaks, they 

share a common ROI. Another feature is the tabulation of the total count for each pertinent ROI. 
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 43 of 177

ROI 3 is the total count for the 60-keV to 63-keV ROI, and ROI 4 is the background count for the 60-

and 63-keV photopeaks. ROI 4 was also used as the count for the 93-keV photopeak, and ROI 5 was 

its background. ROI 2 was probably the count in the L X-ray region, but it was not used.

Figures C-23 and C-24 are examples of the next generation of reports for the germanium detector 

systems. The innovation is the data capture in 10 ROIs. In Figure C-24, the ROIs are labeled with 

the photopeak of interest. Although the data were captured, most of the data were not used, mainly 

because the relationship between the photopeak and its background was not established or was too 

variable [73]. ROI 5 (BKG in Figure C-24) is the common background (divided by a factor) for both 

the 60- and 63-keV photopeaks.

Figure C-25 is an example of a report for a worker with a confirmed deposition. There are no new 

fields.

Figure C-26 is an example of a report for a worker with no detected deposition and illustrates a 

frequent problem with the L X-ray data, namely low-end electronic noise in one or more of the 

detectors. Dose reconstructors should disregard all L X-ray data (including the 13- and 17-keV 

ROIs) [74].

Figure C-27 is an example of a report on which data for the 93-keV photopeak are analyzed and 

presented.

Figure C-28 is an example of the next generation of reports. On this report, the ROI data for each 

detector are tabulated separately, as is the sum. ADC #1 stands for analog-to-digital converter for 

detector #1, which in this case is an EG&G detector, and similarly for the other detectors. This report 

does not report the results in terms of the fraction of the MPLB, an obsolete concept since 1989 [75].

Figure C-29 is an example of the lung count report from an early version of ABACOS-Plus that was 

used through mid-February 1997. Because this software is based on a peak-search method, no ROI 

data are available. In addition, if a uranium or americium peak was not found, the activity was 

reported as less than the decision level [76].

Figure C-30 is an example of the lung count report from ABACOS-Plus after mid-February 1997, 

when the reporting protocol was changed. The primary change was that the activities of 235U, 238U, 

and 241Am are calculated and reported, even if the peak was not detected or if the result was negative. 

The MDA values are for the average worker, as stated on the report. The MDA value for 238U is lower 

than the worker-specific decision level for this case. The worker-specific MDA should be at least 

twice the worker-specific decision level.

Figure C-31 is an example of the lung count report from ABACOS-Plus for a worker with a confirmed 

deposition. The software calculated the deposition for the plutonium isotopes based on the intake 

date in the header and on the calculated ppm 241Am (including ingrowth), which was based on the 

value of the initial ppm 241Am in the worker’s file. The % Error for 241Am was assigned to the 

plutonium isotopes. The basis of the decision level for the plutonium isotopes is not obvious, but was 

probably the decision level for detecting the L X-rays. In any case, this decision level value does not 

apply and should be disregarded for the plutonium isotopes [77]. The value of the ppm 241Am on the 

date of the count was not reported on lung count reports that were generated by ABACOS-Plus. This 

value can be calculated using Equation B-18 in Attachment B and the value of the initial tabulated 

ppm 241Am generally on one of the early lung count reports [78].

Much of the information from ABACOS-Plus is not useful, including Count Rate, Detector Count Rate, 

Analysis Limits, and the total activity.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 44 of 177

Dose reconstructors should note the intake date. If the intake date is different from the date for Count 

Started, the intake date is from the file for a worker with a confirmed deposition. Otherwise, the date 

of the lung count should be used as the intake date [79].

Dose reconstructors should be aware that the lung counter detectors were also used for wound 

counts (Berger 1988b; RFETS 2000b, p. 93). Reports of wound measurements, including the 

calibration of the detector using americium and plutonium sources, look the same as the lung count 

reports except for some header information (name, employer, job code, reason, height, or weight).

It is important to note that the calculated activities for plutonium for lung counts were based on a 

specific, actual, or assumed intake date and initial ppm 241Am. The plutonium values are valid and 

appropriate only for that intake data. If dose reconstructors choose to use another intake date or 

initial ppm 241Am, they should recalculate the set of plutonium lung deposition activities based on the 

recalculated ppm 241Am for ingrowth. This is accomplished by multiplying the original activity of 

plutonium by the ratio of the original ppm 241Am on the date of the count divided by the new value of 

the ppm 241Am on the date of the count. The new value of the ppm 241Am on the date of the count can 

be calculated using Equation B-18 in Attachment B. Dose reconstructors should adjust the activities 

for the discontinuity factors presented in Attachment B. In general, use of the discontinuity factors is 

favorable to claimants [80].

5.7 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

Much of the information in this TBD, including the Attachments, was written by Roger B. Falk and is 

based on his insights, recollections, research and development activities, and administration in the 

radiation dosimetry and health effects programs at the Rocky Flats Plant.

[1] Falk, Roger B. Oak Ridge Associated Universities (ORAU) Team. Senior Life Scientist. June 

2006.

The statements of the primary types of intakes and bioassay data are based on the 

observations by the author during his work at RFP in the internal dosimetry and health effects 

programs.

[2] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is based on the observations of such notations on incident and lung count 

reports related to ZPPR materials.

[3] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The multiplier for WG plutonium is the inverse of Equation B-17 in Attachment B. This 

multiplier is modified to apply to ZPPR plutonium based on the ratio of the weighted specific 

activities of the 239Pu and 240Pu for WG and ZPPR plutonium, 0.071 and 0.0767, respectively. 

The ratio of 0.926 times 48.2 results in the value of 44.6 in the ZPPR multiplying factor.

[4] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Oxides of plutonium metal (air-oxidized and fire-oxidized) are classified as type S and most 

other plutonium compounds as type M by the ICRP (ICRP 1994a). In any case, dose

reconstructors should use the solubility class that is most favorable to claimants.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 45 of 177

[5] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Retention of plutonium in the lungs of workers exposed in the 1965 plutonium fire was 

observed to be more avid than would be predicted by the default ICRP type S model (ICRP 

1994b), based on lung counts performed as part of the Former Radiation Worker Medical 

Surveillance Program at RFP, 28 to 38 years after intake.

[6] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Oxides (air-oxidized and fire-oxidized) of plutonium metal are classified as type S and most 

other plutonium compounds as type M by the ICRP (ICRP 1994a). In any case, dose 

reconstructors should use the absorption type that is most favorable to claimants.

[7] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The recommendation is favorable to claimants when intakes are assessed from airborne 

plutonium data and is essentially neutral when intakes are assessed from urine or lung count 

data.

[8] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The source of the americium is only from the decay of 241Pu. No other americium isotopes are 

involved.

[9] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Spot urine samples for plutonium were rarely requested and then usually associated with a 

significant incident, especially an incident with chelation (DTPA) treatment follow-up. Such 

exceptions should be easily discernible in the documentation of the incident in the worker’s 

health physics file, especially starting in 1990, the period of interest for this recommendation. 

In addition, many of the urine sample result reports include the volume of the sample [see 

Figures C-8 to C-12. (Although some of these examples are for uranium analytes, the format 

of the report is the same for plutonium analytes.)]. These reported volumes can be used to 

normalize the result to a 24-hr sample when appropriate.

[10] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The original 2003 version of this sentence was stated incorrectly. The factor of 1.0264, when 

multiplied by the intake assessed from 239,240Pu urine data, would yield the intake for 

238,239,240Pu, not the 238Pu component of the intake. In addition, the factor was based on a 

slightly different isotopic composition from that stated in Table 5-1. The 238Pu component of 

the intake is obtained correctly by multiplying the intake assessed from 239,240Pu urine data by 

0.0235, a value obtained by dividing the 238Pu fraction of alpha activity stated in Table 5-1 by 

0.98, the sum of the fractions of alpha activity for the 239Pu and 240Pu isotopes.

[11] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Interferences that add to the value of the analyte are always favorable to claimants. 

Therefore, the recommendation was made to use the results as found in the record unless 

dose reconstructors have generic instructions, outside the purview of this TBD, to do 

otherwise.

[12] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is based on the property of the PHA system to separate and count the alphas 

by their energies. The alpha energies of the 239Pu and 240Pu isotopes were sufficiently 

different from the alpha energies of americium and thorium to allow plutonium analyses to be 

unaffected by the presence of americium or thorium, if any.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 46 of 177

[13] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement, based on informal discussions in the 1980s and early 1990s with [name 

redacted], the [position redacted] at Rocky Flats starting in [date redacted], was included here 

for completeness.

[14] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The recommendation for dose reconstructors to be wary of, and generally not use, urine data 

that was flagged with Code 1 is based on good science and common sense. Although 

chelation-enhanced urine data might be favorable to claimants, use of such data without 

modification in standard models that are based on unenhanced data is not scientifically sound. 

It is also not sound to use data that did not pass quality standards in real time.

[15] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The information in this paragraph is based on observations and deductions of the author from 

review of original urine data logs and individual urine data reports in preparation of this TBD 

section and Attachment A, as well as from personal involvement in the development of 

improved urinalysis reporting protocols at RFP in the 1980s and early 1990s.

[16] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is based on an undocumented conversation with [name redacted], [position

redacted] at RFP in the summer of 2003.

[17] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The recommendation to use the values in Table 5-6 seemed to be the only viable option.

[18] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This information is based on the author’s recollections of the implementation of these 

upgrades at RFP.

[19] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is based on the observation that the same reporting format used for plutonium 

results was used for americium results.

[20] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This information is based on informal conversations in the early 1990s with [name redacted], 

the [position redacted] at Rocky Flats starting in [date redacted].

[21] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This credibility check is presented to dose reconstructors as optional guidance and does not 

preclude the use of americium urine data for dose reconstructions if deemed appropriate, even 

if the data do not pass this credibility test. Note also that the maximum ingrowth of americium 

in virgin WG plutonium (with 0.5% by weight 241Pu) is less than 5,000 ppm. Plutonium with 

10,000 ppm 241Am or greater would be credible only for a process that enhanced the 

americium concentrations, such as those processes listed.

[22] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This recommendation was based on the plutonium-to-americium activity ratio, which is 

considerably greater than 2 for WG plutonium, and the problem of the thorium interferences in 

americium urinalyses. These two factors make the plutonium urine data set the better choice 

to determine plutonium intakes.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 47 of 177

[23] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This recommendation extrapolates a value favorable to claimants forward to the next point in 

time for which the reporting level was determined.

[24] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is based on the observation that the same reporting format used for plutonium 

results was used for americium results.

[25] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

If there are interferences that contribute to the magnitude of the results, considering those 

interferences as EU results in a higher than actual outcome and is therefore favorable to 

claimants.

[26] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

It is not clear in the urine data logs for 1964 to 1971 which electroplated uranium samples 

were for EU and which were for DU. EU operations were discontinued in this period although 

some urine sampling for EU could have occurred for workers involved in decontamination 

activities.

[27] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The statements about the units of the reported urine data are based on observations of 

numerous urine data reports.

[28] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This recommendation implies dividing the intake assessed from the assumed 238U urine data 

by 0.89 (the 238U fraction of total DU activity) to calculate the total DU intake. This approach is 

favorable to claimants by about 12% if the activities of the 234U and 235U were actually included 

in the reported uranium urine results in the 1980s.

[29] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This recommendation extrapolates the earlier practice for a period when the actual practice is 

not known. This approach is favorable to claimants if the earlier values would result in a 

higher intake assessment, as in this case.

[30] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

No documented date was found about when gross alpha analyses were discontinued. The 

year 1973 was estimated following a review of HSDS urinalysis reports for a sampling of 

workers previously sampled for gross alpha, with a finding of no analysis code G after 1972.

[31] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Urinalysis code G was observed in HSDS urinalysis reports to 1972. Code G correlates with 

the gross alpha B2 analysis code on the Urinalysis Record Card (see, for example, Figure C-2).

[32] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The term “favorable to claimants” is used interchangeably with the “upper bounds.”

[33] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The effect of interferences is not included in the estimate of the standard deviation because it 

is not really a random variable but rather an intermittent bias of unknown magnitude.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 48 of 177

[34] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Liquid scintillation was used as the counting method for tritium in the 1980s, based on the 

author’s personal observations. No documentation has been noted about the earlier systems 

at RFP, but it seems reasonable to consider that the same method was used in the 1970s.

[35] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Numerous tritium results at these levels have been observed by the author in the worker’s 

health physics files at the cited levels.

[36] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1973 at RFP. Figure C-18 is an 

example of the implementation of these modifications.

[37] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This interference, though troublesome in real time, is favorable to claimants whose americium 

lung count results were enhanced by count from 234Th.

[38] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Interferences, especially contamination on the worker’s chest, occasionally caused a false 

positive lung count. This statement was intended to alert dose reconstructors to this 

possibility. The sentence is modified in recognition of the fact that an intake could have 

occurred without resulting in a deposition in the alveolar-interstitial region of the lung.

[39] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1974 at RFP.

[40] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This modification was implemented by the author in the early 1980s at RFP.

[41] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The cutoff, as defined, is based on limiting the probability of a Type I error (false positive) in 

the signal domain to 5%. Figure C-22 is an example of the implementation of this decision 

level.

[42] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The worker-specific MDA for the americium in vivo measurement depends on the worker’s 

index and the calibration factor K for that index for the detector system used for the worker’s 

lung count. A relatively easy method is to normalize the MDA value for the 1.35 index by the 

ratios of the calibration factors (given in Attachment B for each detector system) for index 1.35, 

and the worker-specific index is:

1 35

worker 1 35

worker

.

.

K MDA MDA

K = (5-1)

An easy method to derive the worker-specific MDA from values listed in Table 5-13 is to use a 

spreadsheet to plot the MDA values for the three indices, for the detector system of interest, 

and to determine the equation for an exponential trend line. This equation, in the form 

y = #.#### e#.#### x

, where x is the worker’s index and y is the worker-specific MDA, can then 

be used to calculate the worker-specific MDA.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 49 of 177

[43] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The recommendation to use the median value of the MDA is consistent with the generic 

approach of the program. Except for some workers with lung counts only in the 1960s, this 

situation is expected to be rare.

[44] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The recommendation to use an initial value of 100 ppm Am, if the actual value for an intake is 

not documented, is based on freshly purified plutonium (within 0 to 5 months depending on the 

efficiency of the purification process).

[45] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This guidance is based on the consideration that significant intakes of plutonium at RFP were 

acute intakes, albeit sometimes a set of intermittent acute intakes. After implementation of the 

body counter in 1965 and as the sensitivity of the system improved, the assignment of the 

intake date to newly detected depositions, but not from a new intake, was problematic. Dose 

reconstructors might have generic instructions, especially for efficiency methods, for assigning 

the intake scenario.

[46] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This guidance is presented to support the calculation of the ingrowth of 241Am after the date of 

an acute intake or after the start of a chronic intake. After the end of plutonium production 

activities in 1989, the initial fraction of 241Pu in RFP plutonium was a decreasing variable 

based on the age of the plutonium since blending.

[47] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1976 and 1981 at RFP. Figures C-20 

and C-21 are examples of the implementation of these modifications.

[48] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The validity of this assumption, listed as an uncertainty, seems to be supported by 

observations that americium lung count measurements for many RFP workers with confirmed 

lung depositions of plutonium-americium mixtures have remained relatively constant or have 

slightly increased at decades after the initial short-term clearance period of several years. 

Recent U.S. Transuranium and Uranium Registries autopsy data for RFP cases also indicate 

the retention of americium in the lungs that is consistent with this assumption.

[49] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

No documented date was found about when gross alpha analyses were discontinued. The 

year 1973 was estimated after a review of HSDS urinalysis reports for a sampling of workers 

previously sampled for gross alpha, with a finding of no analysis code G after 1972.

[50] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Dividing the 238U MDA by 0.89 accounts for the contribution to the DU MDA from activities of 

the other uranium isotopes.

[51] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

If a super-equilibrium situation was operative and the 234Th lung count result was used to 

calculate the DU assuming equilibrium, the calculated DU would be higher than the actual 

activity. Therefore, the approach is favorable to claimants.

[52] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This method suggested to estimate the uncertainty of the activity from its MDA is the same as 

the method suggested in Sections 5.3.1.2, 5.3.3.1.2, and 5.3.4.2 and is equally applicable.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 50 of 177

[53] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Because dose reconstructors are likely to find numerous wound count reports in files of 

workers assigned to plutonium areas, this statement helps to focus the attention of dose 

reconstructors on the most relevant data for quantifying internal doses to organs. The relevant 

data are cited in the next sentence. The actual wound count and contamination data might be 

relevant if the cancer site coincided with the site of the wound, an occurrence not yet noted by 

the author.

[54] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The interpretations are those of the author, either gleaned from using the data as an internal 

dosimetrist at RFP or as the designer and implementer of the reports as part of the technical 

staff supporting RFP internal dosimetry programs.

[55] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The urine data record written on the Urinalysis Record Card preceded the HSDS and was the 

probable source of the urine data loaded into the HSDS and its mainframe database 

predecessors. Because there could have been transcription errors during the preparation of 

the data (punched cards in the 1960s) for loading into the mainframe, the data of the 

Urinalysis Record Cards (the source data) should be considered the correct data, as 

recommended.

[56] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This recommendation is reasonable because the basis for the change is not known and the 

change was not made in the HSDS. It is favorable to claimants because the original record is 

the higher value.

[57] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Exceptions to this practice have been observed in the HSDS urinalysis reports for some 

workers with documented chelation therapy, especially for americium results from analysis of a 

urine sample also analyzed for plutonium (and the plutonium result was correctly coded with 

Code 1).

[58] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This circumstance is evident in the example reports in Figures C-2, C-3, and especially C-5.

[59] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This is presented for information only. The decision of whether to disqualify the result was the 

call of the laboratory quality assurance officer who reviewed the data and signed the report. 

The Batch Val code V is the primary indicator of a valid result.

[60] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This is presented for information only. In practice, the batch evaluation review would have 

occurred before the release of the Analytical Report. The Batch Val code V is the primary 

indicator of a valid result.

[61] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

It is prudent not to use a result that was invalidated based on failure to meet quality standards.

[62] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The original 2003 version of this sentence was stated incorrectly. The factor of 1.0264, when 

multiplied by the intake assessed from 239,240Pu urine data, would yield the intake for 

238,239,240Pu, not the 238Pu component of the intake. In addition, the factor was based on a 

slightly different isotopic composition from that stated in Table 5-1. The 238Pu component of 
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 51 of 177

the intake is obtained correctly by multiplying the intake assessed from 239,240Pu urine data by 

0.0235, a value obtained by dividing the 238Pu fraction of alpha activity stated in Table 5-1 by 

0.98, the sum of the fractions of alpha activity for the 239Pu and 240Pu isotopes.

[63] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The offsite laboratory sometimes was not provided the date on which the worker excreted the 

urine sample. In such cases, the excretion date was written on the report, as was the case for 

the report in Figure C-8. The guidance for dose reconstructors is to use the date written on 

the report if such a situation occurs.

[64] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This is presented for information only. The time of the end of the excretion period is not critical 

for retrospective dose reconstructions. IMBA, which is used by Project dose reconstructors,

has a default sample time of 12:00 a.m.

[65] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The units for fecal and nasal smear sample results are well known to the author from his 

experience as internal dosimetrist at RFP. In addition, the units are probably stated explicitly 

on other reports in the health physics files of affected workers.

[66] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The time of day of the count is not critical for retrospective dose reconstruction. Later lung 

count reports usually recorded the time of day in this field and noted the count time only if it 

was different from the standard count for the era, through the 1970s (see Attachment B for the 

standard count times). Electronically generated lung count reports, starting circa 1981, record 

the count time used for that count (see Figures C-21 to C-31). The count time would be useful 

to dose reconstructors mainly to calculate an MDA for a given lung count, if needed.

[67] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The value of the MPLB for plutonium alpha emitters (239Pu and 240Pu) was calculated using 

Equation 4 in ICRP Publication 2 (ICRP 1959) for an annual dose of 15 rem (0.3 rem/wk), 

organ mass (m) = 1,000 g, f2 = 1, and ε = 53 (from ICRP Publication 2, Table 5, and based on 

a relative biological effectiveness = 10).

[68] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The main reason the measurements obtained by the detector over the gut/liver/below-sternum 

area are not dosimetrically useful is that a calibration factor was not developed in real time to 

convert the signal to activity. It was a subjective measurement; i.e., was it normal or high?

[69] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This ratio was the subjective rule-of-thumb used by the author in real time at RFP as a 

supplemental method to discern possible low-level depositions of the plutonium-americium 

mixtures for lung counts performed with the NaI detector system.

[70] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The variability in the background in the L X-ray region of the spectrum prevented the 

establishment of a stable calibration factor for the direct measurement of plutonium via 

L X-rays. The author was directly involved in this effort.

[71] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

See, for example, Figure C-22.
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[72] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement reflects the practice to perform a follow-up lung count with a better resolution 

detector system (i.e., a germanium detector system) when action levels for a count with a 

poorer resolution, NaI scintillation detector system, such as the phoswich detectors, were 

exceeded.

[73] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

See, for example, Figure C-26.

[74] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Dose reconstructors are advised to disregard these L X-ray data because the counts were 

unreliable because of low-end electronic noise. Because of this unsolved problem in real time, 

a calibration factor to convert from count of plutonium activity was not established.

[75] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

DOE Order 5480.11, implemented in 1989, shifted the basis from the ICRP Publication 2 

approach to control the dose to a critical organ to the ICRP Publication 26 and 30 approach of 

assessing the committed dose equivalent to organs from intakes (ICRP 1959, 1977, 1979; 

DOE 1988). The quality factor for alpha radiation was increased from 10 to 20.

[76] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This statement is the result of direct observation of information in Figure C-29.

[77] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

Not only is the plutonium decision level inoperative because the decision is based on the 

detection of americium, but also the decision level is never operative for follow-up 

measurements of a confirmed deposition – there is no decision to be made. A decision level is 

operative only if the null hypothesis is operative. The null hypothesis is not operative in this 

example.

[78] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

This information was provided just in case dose reconstructors want to determine the value of 

the parts per million of the 241Am used in the calculation.

[79] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

ABACOS-Plus used the date of the count as the default intake date unless an intake date was 

specifically input for the count. This statement should not be interpreted as guidance to dose 

reconstructors to use that default date in dose reconstructions.

[80] Falk, Roger B. ORAU Team. Senior Life Scientist. June 2006.

The only exception to this statement found by the author is the CWT adjustment factor 

(Equation B-4 in Attachment B) for low indices. For indices less than 0.98, the CWT 

adjustment is less than 1.00.

[81] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

D Plant (Building 991) handled “all materials” as a consequence of its function of shipping, 

receiving, and storage of special nuclear and classified materials for RFP, as well as final 

assembly and inspection of plutonium and EU products in the early years. For more 

information, see “Historical American Engineering Record, Rocky Flats Site, Building 991”

(DOE 2011).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 53 of 177

[82] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The method codes are listed at the top of the Urinalysis Record Card (see Figures C-1 to C-3 

in Attachment C). Units, if not listed on the card, were discerned from the urine data logs.

[83] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was discerned from comparing records in the urine data logs with entries on 

Urinalysis Record Cards.

[84] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The tolerance levels were noted as the Working MDL in some early urine data logs. The 

reporting levels were not stated explicitly in the data logs, but rather were discerned from the 

minimum values calculated in the data logs. These minimum values corresponded to 10% of 

the Working MDL.

[85] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This change in the reporting level for the gross alpha results corresponded to the change to 

using plutonium as the default analyte rather than EU. The other changes in this paragraph 

were discerned from the lowest values recorded in the urine data logs.

[86] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This statement was based on the examination of the urinalysis records of a number of workers 

affected by this practice.

[87] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

These reporting levels were discerned from the lowest values recorded in the americium urine 

data logs.

[88] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This is a description of the general method. How and when the volume adjustments were 

made for each analyte and period are discussed later in the document.

[89] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This is a summary of the observations of the recovery determined from the batch spike versus 

a standard recovery value, based on calculations to reproduce the result in the urine data logs. 

Additional discussions are provided for the analytes later in the document.

[90] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The adjustment of the volume in this manner could have occurred earlier. However, urine data 

logs for 1955 to 1959 were not found.

[91] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was provided by [name redacted], the [position redacted] at Rocky Flats 

starting in [date redacted], in an interview with the author in 1992.

[92] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

Notations indicating detectors with 40% efficiency started to appear in the urine data logs in 

August 1964.

[93] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This statement is based on the author’s direct experience and on discussions with [name 

redacted], who was the [position redacted] at Rocky Flats starting in [date redacted] and also 

[position redacted] during the cited period.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 54 of 177

[94] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The author was directly involved with making this change.

[95] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The author was directly involved with a committee of Radiological Health and Analytical 

Laboratory personnel in 1993 to implement these changes. Upon further review, the author 

observed that the count time was increased to 1,440 minutes (24 hours) rather than the stated 

2,000 minutes. This correction is now made.

[96] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was provided by [name redacted], the [position redacted] at Rocky Flats 

starting in [date redacted], in an interview with the author in 1992; it was verified during 

examination of the urine data logs for the early 1960s.

[97] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

These values were based on observations made in the urine data logs concerning when the 

count results were converted to the activity of EU in the sample. Apparently, that decision was 

based on the count uncorrected by volume, for which the minimum reported activity was 

20 dpm/24-hr sample. When a volume adjustment was made, higher minimum reported 

values up to 28 dpm/24-hr sample were observed.

[98] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

These values were obtained through calculations by the author to duplicate the results stated 

in the urine data logs.

[99] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The urine data logs for 1964 to 1971 do not distinguish explicitly which samples were for 

workers in EU areas versus those for workers in DU areas.

[100] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was discerned by the author from examination of the urine data logs for 

fluorimetric measurements.

[101] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was discerned by the author from examination of the urine data logs for 

electroplating measurements.

[102] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was provided by [name redacted], the [position redacted] at Rocky Flats 

starting in [date redacted], in an interview with the author in 1992.

[103] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

A better way to indicate the generosity of the nonspecificity of the gross alpha result if applied 

to a specific radionuclide is to use the term “favorable to claimants” rather than “upper 

bounds.”

[104] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

To have a coherent data set, only background count data for samples counted for 150 minutes

were extracted from the urine data logs.

[105] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The composite value was used because the detector background appeared to be reasonably 

stable in the 1950s and 1960s, as observed in the previous table.
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[106] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This exception was made for americium because the detector backgrounds for the 1950s did 

not apply.

[107] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This statement is a summary of the observations of the author during the review of the urine 

data logs.

[108] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This approach is consistent with the consideration stated in the fourth bullet in the subsection 

headed Assessment of MDA.

[109] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This subjective observation of the similarities of the recovery values in the preceding table was 

interesting to the author but was not used to determine any MDA value.

[110] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The distribution of volumes for routine 24-hour urine samples was determined from the 

volumes recorded in the urine data logs for gross alpha analyses for 1967 and 1971, a data 

set of 1,437 values. The author chose the gross alpha samples as the sample set least likely 

to include special samples that could have had an excretion period of less than 24 hours.

[111] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The distribution of volumes for routine 24-hour urine samples was determined from the 

volumes recorded in the urine data logs for gross alpha analyses for 1967 and 1971, a data 

set of 1,437 values. The author chose the gross alpha samples as the sample set least 

probably to include special samples that could have had an excretion period of less than 

24 hours.

[112] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

Electrodeposited plutonium and americium samples were marked in the data logs with an E. 

No similar designation has been observed by the author in any reports of these urinalysis 

results.

[113] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

No evidence of a systematic bias in the background or the calibration factor was discerned by 

the author. Therefore, ΔB and ΔK were set equal to zero.

[114] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This information was discerned from the urinalysis data logs by the author.

[115] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The decision by the author to assess the MDA based on one aliquant was based on the 

observation of the data logs that the decision of detection for the overwhelming majority of the 

samples was based on only one aliquant. Occasionally, the decision was based on the 

average of two aliquants. Because the MDA for one aliquant is higher than that for two 

aliquants, this decision is consistent with the consideration stated at the beginning of the 

Assessment of MDA subsection, fourth bullet.

[116] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

This introduction summarizes the information presented in more detail in the body of this TBD. 

Most of this information is based on the direct experience of the author, who provided 

technical support to the operations and developments of the in vivo lung-counting systems at 
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 56 of 177

RFP starting in 1970 and extending into the mid-1980s and also from 1989 to 1992. The 

author also provided second-level management of the dosimetry programs from 1986 to 1989.

[117] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The dates for the start of routine operations of the three counting rooms were determined from 

the references [Room A (Boss and Mann 1967) and Room C (Falk et al. 1979)] or from the 

author discerning the year that the room was first recorded on a worker’s lung count report 

(Room B).

[118] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The information summarized in this section was discerned by the author from (1) notations on 

lung count reports for workers counted in that era, (2) a transition briefing from [name

redacted], the [position redacted] providing technical support to the Rocky Flats [position

redacted] from [date redacted] to [date redacted], and (3) reports cited in the reference 

section.

[119] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The information in this section describes the lung-counting program as found by the author in 

1970. The transitions in 1969 were verified by the author in preparation of this report by 

observations of notations in worker lung count reports. The transitions did not generally take 

place exactly on January 1, 1969, but usually sometime in 1968, which was a transition year. 

However, for the purpose of determining MDAs, the old practice was considered to be 

extended through the year, and no credit was taken for the new practice until it was in effect 

for the entire year.

[120] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The date of the first entry in the logbook for routine ppm 241Am determinations for incident 

samples was observed by the author to be January 3, 1969. This statement does not preclude 

earlier special ppm 241Am determinations. Indeed, a special ppm 241Am determination was 

made for the October 15, 1965, plutonium fire incident.

[121] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The information in this section describes the systems based on the direct involvement of the 

author and verified by the author from observations of the body count result reports as 

needed.

[122] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The information in this section describes the systems as observed by the author, either directly 

or indirectly.

[123] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The upgrades in this period were implemented by [name redacted] and [name redacted], 

technical staff supporting the [position redacted] program at Rocky Flats in the [date redacted]

into the [date redacted], with discussions with the author, who at that time was a customer of 

the in vivo measurements program as the internal dosimetrist for the medical monitoring 

program for former radiation workers at RFP. This era is well documented in RFETS (2000b).

[124] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The count time, which is the duration of the lung count, was observed by the author from lung 

count reports for workers in this era.
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[125] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

For this equation and for subsequent equations of the calibration factor K, the calibration 

factors were normalized to LLNL phantom [also called the Lawrence Livermore Torso 

Phantom and described in RFETS (2000b)]. Normalizing to this phantom is consistent with 

the approach described in the second paragraph in Section B.4 of Attachment B because this 

phantom was used in “the most recent calibration method.”

[126] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

See the subsection headed 1969 to 1976 in Section B.3. In addition, the information in this 

section describes the lung-counting program as found by the author in 1970. The transitions 

in 1969 were verified by the author in preparation of this report by observations of notations in 

worker lung count reports. The transitions did not generally take place exactly on January 1, 

1969, but usually sometime in 1968, which was a transition year. However, for the purpose of 

determining MDAs, the old practice was considered to be extended through the year, and no 

credit was taken for the new practice until it was in effect for the entire year.

[127] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The author personally determined the resolution of the Ortec and Princeton Gamma Tech 

(PGT) detectors for the 59.5-keV photopeak of the 241Am gamma and noted the degradation in 

the resolution. This was expected because the collection efficiency of the charge induced in 

the active part of the detector diminishes with an increasing volume of that active part (based 

on the author’s recollection of solid-state physics).

[128] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The author personally determined the resolution of the Ortec and Princeton Gamma Tech 

(PGT) detectors for the 59.5-keV photopeak of the 241Am gamma and noted the degradation in 

the resolution. This was expected because the collection efficiency of the charge induced in 

the active part of the detector diminishes with an increasing volume of that active part (based 

on the author’s recollection of solid-state physics).

[129] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

See Table 5-1 in the body of this TBD. The value of 0.0049 had been historically used at the 

RFP body counter as the rounded value of 0.005.

[130] Falk, Roger B. ORAU Team. Senior Life Scientist. July 2006.

The method of determining the ppm 241Am from samples representative of the plutonium 

mixtures involved in possible inhalation incidents, starting in 1969, involved the ratio of the 

L X-ray photopeaks and the 241Am 59.5-keV photopeak as measured by a NaI(Tl) detector. 

The ppm 241Am determined by this method was highly uncertain for values less than 100 ppm 

and greater than 10,000 ppm because of the counting statistics. Although one might consider 

zero ppm 241Am to be the true lower bound for freshly purified plutonium, a zero value is not 

practical to use in Equation B-17 (i.e., division by zero is not allowed). The value of 100 ppm 

241Am is also supported by its rank at the 10th percentile in the low-to-high ranking of 442 

values of the incident ppm recorded in the logbook for January 1969 to September 1972. The 

value also represents the ingrowth of americium in freshly purified plutonium (within 0 to 

5 months, depending on the efficiency of the purification process).

[131] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

Lognormal distributions typically provide the best fit to the available data and are a distribution 

suitable for input into the Interactive RadioEpidemiological Program (IREP).
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[132] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

The error associated with individual bioassay results is normally distributed because the 

dominant source of uncertainty is the counting statistics. Although the underlying group 

statistics are normally distributed, each result was treated as if it were normally distributed to 

match what is done for analysis of an individual’s bioassay data and because the lognormal 

distribution of the data is addressed by analyzing both the 50th- and 84th-percentiles of the 

data.

[133] Arno, Matthew. ORAU Team. Dose Reconstructor. June 18, 2007.

The use of the 95th-percentile intake value was required as part of the resolution of SEC 

Petition SEC-00030 for Rocky Flats (NIOSH 2006a).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 59 of 177

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Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 21.

NIOSH (National Institute for Occupational Safety and Health), 2006a SEC Petition Evaluation Report, 

Petition SEC-00030, National Institute for Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, April 7. [SRDB Ref ID: 132777]

NIOSH (National Institute for Occupational Safety and Health), 2006b, Rocky Flats Internal Dosimetry 

Coworker Extension, OCAS-TIB-014, Rev. 0, Office of Compensation Analysis and Support,

Cincinnati, Ohio, December 7. [SRDB Ref ID: 29935]

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5.

NIOSH (National Institute for Occupational Safety and Health), 2011, Technical Basis Document: Site 

Profiles for Atomic Weapons Employers That Worked Uranium Metals, Battelle-TBD-6000, 

Rev. 1, Office of Compensation Analysis and Support, Cincinnati, Ohio, June 17.

NIOSH (National Institute for Occupational Safety and Health), 2013 SEC Petition Evaluation Report, 

Petition SEC-00192, National Institute for Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, September 30. [SRDB Ref ID: 

132776]

ORAUT (Oak Ridge Associated Universities Team), 2004, Coworker Data Exposure Profile 

Development, ORAUT-PLAN-0014, Rev. 00, Oak Ridge, Tennessee, November 24.
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ORAUT (Oak Ridge Associated Universities Team), 2005, Technical Information Bulletin: Guidance 

on Wound Modeling for Internal Dose Reconstruction, ORAUT-OTIB-0022, Rev. 00, Oak 

Ridge, Tennessee, November 18.

ORAUT (Oak Ridge Associated Universities Team), 2009, Technical Information Bulletin: Use of 

Claimant Datasets for Coworker Modeling, Rev. 00, ORAUT-OTIB-0075, Oak Ridge, 

Tennessee, May 25.

ORAUT (Oak Ridge Associated Universities Team), 2010, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-2, Oak Ridge, Tennessee, November 

29.

ORAUT (Oak Ridge Associated Universities Team), 2012a, untitled letter to J. Gueretta (U.S. 

Department of Energy, Office of Legacy Management) with collection of dosimetry records for 

[Name 1 redacted], Cincinnati, Ohio, May 14. [SRDB Ref ID: 113622]

ORAUT (Oak Ridge Associated Universities Team), 2012b, untitled letter to J. Gueretta (U.S. 

Department of Energy, Office of Legacy Management) with collection of dosimetry records for 

[Name 2 redacted], Cincinnati, Ohio, May 14. [SRDB Ref ID: 113623]

ORAUT (Oak Ridge Associated Universities Team), 2012c, untitled letter to J. Gueretta (U.S. 

Department of Energy, Office of Legacy Management) with collection of dosimetry records for 

[Name 3 redacted], Cincinnati, Ohio, May 14. [SRDB Ref ID: 113624]

Ray, E. L., 1953, “Monthly Progress Reports for Personnel Meters, Body Counts, and Bioassay Data, 

1953–1963,” memorandum to C. W. Piltingarud, The Dow Chemical Company, Golden, 

Colorado, November 4. [SRDB Ref ID: 24631]

RFETS (Rocky Flats Environmental Technology Site), 1981, Neptunium Processing at the Rocky 

Flats Plant, RFP-2899, Rocky Flats Plant, Golden, Colorado, September 20. [SRDB Ref ID: 

24722]

RFETS (Rocky Flats Environmental Technology Site), 1992, ALPHA QCBLK2.LIS;1, Computer Data 

Report 16-JAN-1992, U.S. Department of Energy, Golden, Colorado. [SRDB Ref ID: 23789]

RFETS (Rocky Flats Environmental Technology Site), 1998a, Internal Dosimetry Technical Basis 

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RFETS (Rocky Flats Environmental Technology Site), 1998b, Statement of Work for Analytical 

Measurements, General Bioassay Services, Module BA01-B.4, Golden, Colorado, May 20. 

[SRDB Ref ID: 8800]

RFETS (Rocky Flats Environmental Technology Site), 1998c, Internal Dosimetry Technical Basis 

Documentation, Other Isotopes, Section 7, Version 3, Golden, Colorado, April 1. [SRDB Ref 

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RFETS (Rocky Flats Environmental Technology Site), 1998d, Internal Dosimetry Technical Basis 

Documentation, In-Vitro Radiobioassay, Section 2, Version 3, Golden, Colorado, April 1. 

[SRDB Ref ID: 8798, p. 59]
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RFETS (Rocky Flats Environmental Technology Site), 2000a, Report on the Flow of Recycled 

Uranium at the Department of Energy’s Rocky Flats Plant, 1953–1993, U.S. Department of 

Energy, Golden, Colorado, June 30. [SRDB Ref ID: 16227]

RFETS (Rocky Flats Environmental Technology Site), 2000b, Internal Dosimetry Technical Basis 

Documentation, Section 3, In-Vivo Radiobioassay, Version 4, Golden, Colorado, November 

27. [SRDB Ref ID: 8798, p. 69]

RFETS (Rocky Flats Environmental Technology Site), 2002, Internal Dosimetry Technical Basis 

Documentation, Section 5, Plutonium and Americium, Version 6, Golden, Colorado, August 

30. [SRDB Ref ID: 8798, p. 117]

Ulsh, B. A., B. L. Rich, M. H. Chew, R. L. Morris, M. Sharfi, and M. R. Rolfes, 2008, “Establishing 

Bounding Internal Dose Estimates for Thorium Activities at Rocky Flats,” Health Physics, 

volume 95, number 1. [SRDB Ref ID: 122613]

Vejvoda, E., 2005, Summary of Rocky Flats Plant Waste Buried in the Subsurface Disposal Area, 

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[SRDB Ref ID: 33009]
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GLOSSARY

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure.

alpha particles

See alpha radiation.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure.

curie

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

detection limit (lower)

See limit of detection.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rad, rep, or grays.

dosimetry

Measurement and calculation of internal and external radiation doses.

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 

some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. For 

example, the isotope of uranium that contains 92 protons and 143 neutrons can appear as 

235U, U-235, or uranium-235. See periodic table of the elements and radioactive isotope.

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.
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extremities

Portion of the arm from and including the elbow through the fingertips and the portion of the 

leg from and including the knee and patella through the toes.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

gamma ray

See gamma radiation.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element.

limit of detection

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level. See minimum detectable level.

maximum permissible lung burden (MPLB)

Historical occupational limit on the amount of a radionuclide present in the systemic body at 

the end of 50 years as a result of being exposed at the maximum permissible concentration for 

50 working years.

minimum detectable activity (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.
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nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

periodic table of the elements

Arrangement of the chemical elements in order of increasing atomic number from left to right 

and by similar chemical properties vertically. Elements of similar properties occur one under 

the other, which yields groups or families of elements.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

radioactive isotope

Natural or synthetic form of an atom that emits radioactivity when it decays. See isotope.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radioactive

Of, caused by, or exhibiting radioactivity.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

radionuclide

Radioactive nuclide. See radioactive and nuclide.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

site returns

At Rocky Flats, weapons components returned from other sites for disassembly and recovery 

of materials.

whole-body (WB) dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.
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X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Introduction ............................................................................................................................ 69

A.2 MDA Methodology.................................................................................................................. 69

A.3 History of Methods ................................................................................................................. 70

A.4 Assessment of MDA............................................................................................................... 75

A.5 MDA Values........................................................................................................................... 80

A.6 Discussion ............................................................................................................................. 85

LIST OF TABLES

TABLE TITLE PAGE

A-1 Method codes ........................................................................................................................70

A-2 Correlation of method code and analyte................................................................................. 71

A-3 Values of tolerance and reporting levels................................................................................. 71

A-4 Sample volumes for routine 24-hour urine samples................................................................ 76

A-5 Detector background for gas flow proportional counters......................................................... 76

A-6 Median and 95th-percentile blank count rates........................................................................ 77

A-7 Efficiencies of alpha-counting detectors ................................................................................. 78

A-8 Recoveries used in MDA assessments .................................................................................. 78

A-9 Fraction of alphas absorbed in residue................................................................................... 79

A-10 Gross alpha calibration factor................................................................................................. 80

A-11 Values of variables and MDA for plutonium for median conditions ......................................... 81

A-12 Values of variables and MDA for plutonium for extreme conditions ........................................ 81

A-13 MDA for plutonium for one, two or three extreme conditions .................................................. 81

A-14 Values of variables and MDA for EU for median conditions.................................................... 82

A-15 Values of variables and MDA for EU for extreme conditions .................................................. 82

A-16 MDA for EU for one, two, or three extreme conditions............................................................ 83

A-17 Values of variables and MDA for fluorimetric measurements of DU for median and 

extreme conditions................................................................................................................. 83

A-18 Values of variables and MDA for americium for median conditions ........................................ 84

A-19 Values of variables and MDA for americium for extreme conditions ....................................... 84

A-20 Values of the MDA for americium for one, two, or three extreme conditions........................... 84

A-21 Values of variables and MDA for gross alpha measurements for median conditions .............. 85

A-22 Values of variables and MDA for gross alpha measurements for extreme conditions............. 85

A-23 Values of the MDA for gross alpha measurements for one, two, or three extreme 

conditions...............................................................................................................................86
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ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS (continued)

A.1 INTRODUCTION

Urinalysis was used at RFP since the start of operations in 1952 to detect intakes of radionuclides by 

workers who were exposed, or had the potential to be exposed, to plutonium, EU, or DU. Urinalysis 

involved the submission of a urine sample by the worker, a chemical processing of the sample to 

isolate the radionuclide of interest (the analyte), and measurement and calculation of the mass or 

activity of the analyte in the sample. The request for submission of the urine was either scheduled as 

part of a routine monitoring program or was specially requested after an actual or suspected intake. 

Routine urine samples were typically 24-hour excretions, either one continuous 24-hour period (but 

not taken at the RFP site) or two 12-hour periods. Special urine samples could be 24-hour samples, 

overnight samples, or a single voiding. The chemical processing of the sample depended on the 

analyte and the need for specificity and recovery. “Specificity” refers to separation of the desired 

radionuclide from interferences such as other radionuclides. “Recovery” refers to isolating as much of 

the analyte as possible in the final medium to be measured (counted). The measurement of the 

sample typically involved counting the alpha radiation from the processed aliquant of the sample and 

determining the activity of the analyte in the original sample. Also involved was the fluorometric 

measurement of mass of DU. The assessment of the MDA involves the determination of the activity 

of the analyte in the original urine sample that would be expected to be detected by the methods and 

systems used at RFP. The analytes of interest are plutonium, americium, EU, and DU. In addition, 

RFP analyzed for gross alpha using a nonspecific analysis for workers from 1952 to 1971 who were 

potentially exposed to any of the analytes of interest. This attachment focuses on the period from 

1952 to 1971, for which many of the urinalysis logs have been located and analyzed to obtain the 

information necessary to assess the MDA. This also is the period when urinalysis procedures were 

primitive and evolving and numerous dosimetrically interesting events and intakes were occurring at 

RFP.

A.2 MDA METHODOLOGY

The general equation for the MDA is Equation 6 in the American National Standard, Performance 

Criteria for Radiobioassay (HPS 1996):

(1 2 23 Δ Δ B ks K B )( o ) MDA

KT

+ ++ = (A-1)

where:

B = the total count of the appropriate blank

ΔB = the maximum expected fractional systematic error bound in the appropriate blank

K = calibration factor

ΔK = the maximum fractional systematic error bound in the calibration factor K

k = the abscissa of the standardized normal distribution corresponding to the 0.05 

probability level (for α = 0.05 and β = 0.05, k = 1.645)

so = the standard deviation in the net count of a sample with no additional analyte:

2

2 B0

o B1 2

s

s s

m

= + (A-2)

T = the standard counting time for the procedure
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ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS (continued)

where:

sB1 = the standard deviation of the sample, where the sample contains no actual 

analyte above that of the appropriate blank

sB0 = the standard deviation in the unadjusted count of the appropriate blank

m = the adjustment factor for the appropriate blank

Applying this equation to urinalysis methods at RFP involves determining the value of each variable 

for measurements of the analytes (plutonium, americium, EU, DU, and gross alpha) as the methods 

evolved.

A.3 HISTORY OF METHODS

General Information

In the beginning of operations (1952), RFP was divided into four distinct subplants plus a general 

support area. The subplants were named A Plant, B Plant, C Plant, and D Plant. The designations A, 

B, C, and D are significant because they are also the code names for the materials that were 

processed in those plants as well for the urinalysis procedures that were used to analyze those 

materials. The records of the 1950s do not contain the words “depleted uranium,” “enriched uranium,” 

and “plutonium.” Instead, DU is A material processed in A Plant (buildings numbered 4##, mainly 

Building 444); EU is B material processed in B Plant (buildings numbered 8##, mainly Building 881); 

and plutonium is C material processed in C Plant (buildings numbered 7##, mainly Building 771). D 

Plant (buildings numbered 9##, mainly Building 991) handled all materials [81]. A nonspecific gross 

alpha urinalysis method was used for workers in D Plant. [Note: Building numbers were two-digit 

numbers until 1968, when the numbers were expanded to three digits (e.g., Building 771 was 

originally Building 71)] From 1962 to 1963, the EU operations were phased out at RFP, although 

urinalysis monitoring for EU continued through 1971.

The Urinalysis Record Card is an important and significant record for the early (1952 to 1969) urine 

data and for the methods that generated those data for a specific worker. A Urinalysis Record Card 

was established for each monitored worker and included the result of each urine sample, the date of 

the sample, and the code of the urinalysis method that was used to generate that result are recorded. 

The card is now in the worker’s Health Physics file, which is the primary RFP record of dosimetry 

information for a worker. Table A-1 lists the method codes [82].

Table A-1. Method codes.

Code Meaning

A Fluorimeter, reported in μg/L 1952–1956 and μg/24 hr 1957–1964

B1 Electroplating, reported in dpm/24 hr. (Note: Electroplating, in RFP records, more properly should be 

called electrodeposition.)

B2 Ether extraction, reported in dpm/24 hr

B3 TBP extraction (hand-written on some cards)

C1 Carrier precipitation, reported in dpm/24 hr

C2 TTA extraction, reported in dpm/24 hr. (Note: On the header of cards for 1961–1965, the code C2 is 

“Pu by Radio Autography.” There is no indication that this method was implemented at RFP.)

D TBP extraction

Although there is some correlation of the codes with the subplants, there are some exceptions. 

Table A-2 summarizes the correlation of the method code and the analyte [83].
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ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS (continued)

Table A-2. Correlation of method code and analyte.

Analyte Method code

DU A, B1 (starting 5/1/64)

EU B1

Plutonium C1, C2

Gross alpha B2, B3, D

Tolerance levels were used at RFP in the 1950s and 1960s as an indicator of the maximum 

permissible amount (activity) of a radionuclide excreted per day in a worker’s urine. The technical 

basis for the values of tolerance levels has not been identified. The significance is that urinalysis 

results less than 10% of the tolerance level were recorded and reported as background (BK on the 

Urinalysis Record Card) or zero, regardless of the underlying sensitivity of the method, with some 

exceptions. Table A-3 lists the values of the tolerance levels [84]:

Table A-3. Values of tolerance and reporting levels

Analyte Tolerance level Reporting level

DU 58 μg/24 hr ≥5.8 μg/24 hr

EU 88 dpm/24 hr ≥8.8 dpm/24 hr

Plutonium 8.8 dpm/24 hr ≥0.88 dpm/24 hr

Gross alpha 88 dpm/24 hr ≥8.8 dpm/24 hr

These reporting (and recording) levels continued through April 1964 for both DU and EU, through 

1961 for plutonium, and through 1963 for gross alpha. From May 1964 through 1971, the reporting 

level for DU and EU was ≥20 to 28 dpm/24 hr. After 1963, the reporting level for gross alpha was 

≥0.9 dpm/24 hr [85].

For plutonium, the reporting and recording level was ≥0.2 dpm/24 hr for 1962 to April 6, 1970. After 

that date, all results ≥0.00 dpm/24 hr were recorded and reported. Negative values were recorded 

and reported as 0.00 dpm/24 hr. A further exception is that, for some workers, the practice 

implemented on April 7, 1970, was applied retroactively for their plutonium data. This retroactive 

application was variable depending on how far back it was applied [86].

In 1963, a specific analysis for 241Am was implemented. The recording and reporting level for 241Am 

was ≥0.24 dpm/24 hr in 1963, ≥0.2 dpm/24 hr from 1964 to 1967, and ≥0.3 dpm/24 hr from 1968 to 

1971 [87].

The general method for data analysis for alpha-counting procedures (1952 to 1971) was:

( ) ( Det Blk )( ) dpm/24-hr sample

C V

T A B B

Activity εR

− − = (A-3)

where:

C = Total count

T = Count time (min)

BDet = Detector background count rate (cpm)

BBlk = Reagent blank count rate (cpm)

V = Sample (or standard) volume (mL)

A = Volume of the aliquant analyzed (or volume of the sample, if the entire sample was 

analyzed) (mL)
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MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS (continued)

ε = Efficiency (geometry) of the detector (cpm/dpm)

R = Recovery, fraction of the analyte in the aliquant or sample that is transferred to the 

planchet or disk to be counted

The detector background count rate was generally tabulated in the urinalysis data logs through 1961. 

After 1961, the value for the detector background is implicit in the data reduction but is not explicitly 

recorded. The same detectors were used for alpha counting for all analytes.

Reagent blanks were generally processed with each batch of samples, and the value of the blank 

count rate that was used in the data reduction was generally tabulated in the urinalysis data logs.

The ratio V/A is a volume adjustment factor that was used for two purposes. If the entire sample was 

not analyzed, this ratio normalized the result from the volume of the analyzed aliquant to the total 

sample. If the volume of the total sample was less than a minimum specified volume (e.g., 1,000 mL), 

the sample was considered to be less than a 24-hr sample, and the ratio was used to normalize the 

sample result to that for a 24-hr sample [88]. The sample volume was recorded in the urinalysis data 

log for each sample.

The value of ε was the geometry rating of the detector. In 1952 and 1953, ε was 0.45. After that, the 

detectors were called 50% detectors, and ε was 0.50. In 1964, 40% detectors (ε = 0.40) were added 

to the system as a supplement to the 50% detectors.

The value of R was generally a standard value. Depending on the process, spiked samples (samples 

to which a known activity of the analyte was added) were generally processed with each batch of 

samples. The recovery values that were calculated from the spiked samples were the ratios of the 

count rate of spiked sample to the average count rate of four to six samples deposited on the planchet 

or plate with minimal processing. The recovery values for the spiked samples were not normalized to 

the deposited activity (dpm). In addition, the recovery values from the spikes usually were not used to 

customize the standard value of R for samples in the batch [89].

The fraction of absorption of the alpha particles in the residue on the planchet or plate was not 

explicitly incorporated either in the efficiency or recovery.

The term εR was frequently combined, especially in the 1950s. In the 1960s, the term 1/εR was 

occasionally tabulated in the urinalysis data logs as “R.F.” (presumably for “recovery factor”), and was 

used as a multiplier to convert the net count per minute to activity in the sample.

The general method for the mass measurements of uranium using the fluorimeter (1953 to 1964) was:

( ) Blk μg/24-hr sample S B Mass

K

− = (A-4)

where:

S = Signal reading of the sample aliquant

BBlk = Signal reading of the blank

K = Constant/V [Constant is custom to each process; V = volume (mL) of the entire urine 

sample. If the sample volume ≤1,000 mL, V = 1,000 mL.]
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ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS (continued)

The history of these urinalysis methods is largely based on an interview with the [position redacted]

from [date redacted] to [date redacted], [name redacted], in 1992 and on a review of the bioassay 

data logs from 1952 to 1971.

Plutonium

1952 to 1961

The urine sample was processed using a method called carrier precipitation (also called 

coprecipitation). The plutonium in the urine sample (plus some americium and thorium) was carried 

into the precipitate with lanthanum fluoride. The precipitate was dissolved and the solution was 

evaporated on a planchet, which was counted with a gas-flow proportional counter. Typical count 

time was 150 minutes. A spike sample and a reagent blank sample were processed with the worker 

samples, sometimes with each batch and sometimes less frequently. The result of the spike sample 

might have been used to establish the value of the recovery of the analyte for the batch. Similarly, the 

result of the blank (counts per minute) might have been used to establish the value of the blank 

subtracted from the total count rate of the sample. Detector efficiency was stated to be 0.50. A 

volume adjustment factor (1,200/sample volume) was applied as a multiplier to the result if the sample 

volume was less than 1,000 mL. The first evidence of the use of this factor is in 1960 [90].

1961 to 1962

Starting on December 13, 1961, a thenoyltrifluroacetone (TTA) extraction step was added to the 

carrier precipitation method to improve the specificity of the process to isolate plutonium [91]. No 

other changes were made to the previous method.

1963 to 1978

The ion exchange method replaced the carrier precipitation/TTA extraction method in 1963 and was 

used, with refinements, thereafter. The method was specific to plutonium. In addition, americium 

could be recovered separately from the plutonium in the same sample. Evaporation of the analyte on 

a planchet was continued, but that method was gradually phased out and replaced by 

electrodeposition on a stainless-steel disk. About one-third of the samples were electrodeposited in 

1964 and one-half or more from 1967 to 1971. In 1973, an alpha PHA counting system with surface 

barrier detectors was started with four detectors. The practice of using internal tracers (236Pu or 242Pu) 

for some plutonium samples was begun concurrently. A batch blank continued to be processed, 

although its use was inconsistent. For example, in 1971, a blank count rate of 0.00 cpm was used 

even though the median value of the batch blank was 0.06 cpm. In 1964, detectors with an efficiency 

of 0.4 were used as a supplement to the detectors with 0.5 efficiency [92].

1978 to 1993

By 1978, all counting systems had been converted to the PHA system, and all plutonium samples 

were processed with internal tracers. The recovered fraction of the internal tracer for that sample was 

applied in the analysis of the result for that sample. The acceptable range of the fractional tracer 

recovery was 0.10 to 1.10. The result of a sample was invalidated if the recovery was outside the 

acceptable range [93]. In 1990, the acceptable recovery range was changed to 0.35 to 1.10 [94]. The 

count time of 720 minutes was used for all samples. A batch blank continued to be processed and 

generally was used in the data analysis unless suspected to have been contaminated excessively (a 

subjective decision). In 1985, the blank method was modified. The value of the blank that was used 

in the analysis of the result for a sample was the average value of the last 20 valid batch blanks. To 

be valid, a batch blank value was tested using the Dixon outlier test and, if it passed the test, was 

added to the population of the last 20 blanks. In 1988, the blank process was further modified by use 

of the Winsorized trimmed mean of the population of 20 blanks instead of the average value. The 
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purpose of these modifications was to minimize the influence of laboratory contamination artifacts, 

which were considered to be nonrandom events that, if incorporated in the blank, would 

inappropriately bias the results of the other samples on the low side. In addition, the reagent blank 

was replaced by a matrix blank, either real or artificial urine. The volume of the analyzed sample 

(aliquant) was 800 mL if the volume of the sample was greater than 800 mL or, if the volume of the 

sample was less than 800 mL, the entire sample. The result of the aliquant was divided by the 

volume fraction (800 mL/volume of the sample) if the volume of the sample was ≥800 mL. The 

efficiency of the detectors was typically in the range of 0.25 to 0.35.

1993 and beyond

Upgrades to procedures occurred in 1993 to achieve a process MDA less than or equal to 0.020 

dpm/sample [95]. Count time was increased to 1,400 minutes. The entire sample was analyzed so 

that the volume fraction was unity for all samples. In addition, a contract was established with a 

commercial bioassay laboratory, with a requirement that an MDA of ≤0.02 dpm/sample be achieved. 

In 1997, the onsite bioassay laboratory was shut down.

Americium (1963 and beyond)

Except for the details of the chemistry, the process for americium was similar to that for plutonium. A 

solvent extraction process, specific for americium, was first used in 1963 [96]. A new process (not 

defined in the data log) was started in November 1965. At some point, not defined in the examined 

data logs, the ion exchange method was implemented for americium.

Enriched Uranium (1952 to 1971)

Urine samples were analyzed for EU according to a process called electroplating. A 50-mL aliquant 

of urine was extracted from the 24-hour sample and chemically processed to minimize impurities. 

The resulting solution was poured into an electrodeposition column, and the uranium was deposited 

on a stainless-steel disk. The disk then was counted for alpha radiation with the gas-flow proportional 

counters, as described for plutonium. Counting times in this period were 30, 40, 60, 70, 90, 120, and 

150 minutes.

From 1952 to 1955, one aliquant per sample was used. In 1960, a second aliquant was processed if 

the result of the first aliquant was ≥7 dpm/24-hr sample. If the second result was within a specified 

range of the first result, the average of the two results was recorded and reported. If the second result 

was out of the specified range, a third aliquant was processed, and the average of the two results that 

best confirmed each other was used. If that average was less than the reporting level of 

8.8 dpm/24 hr, the result was recorded and reported as background. From 1961 to 1971, two 

aliquants were routinely processed for each urine sample, with a third aliquant (1961 to 1969) 

processed if the spread of the results of the first set was outside the specified range. The recording 

and reporting logic was the same as that for 1960. From 1964 to 1971, the recording and reporting 

limit appears to have been ≥20 to 28 dpm/24-hr sample, depending on the volume of the sample [97].

Blank data were not used to adjust the sample count rate except sporadically in 1963 and 1964. 

Detector background was usually subtracted, but not always. Spike samples were processed, 

although it is not obvious how those data were used, if at all. Instead, a constant value of the product 

of the detector efficiency ε and the recovery R was used: 0.40 (1953 to 1955 and 1971), 0.30 (1960 

to 1970), and 0.24 (1964 to 1970 for detectors with ε = 0.40) [98].

EU operations were phased out at RFP from 1962 to 1963, although some workers were still 

monitored for EU intakes through 1971 [99].
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Depleted Uranium (1952 to 1971)

Two methods were used to analyze urine samples for DU. From 1952 to April 1964, a fluorimeter was 

used to measure the mass (micrograms) of uranium in a 100λ (0.1-mL) aliquant of the 24-hour urine 

sample. The result was extrapolated to the total sample and reported in the unit of μg/24-hr sample. 

A volume adjustment was made if the sample volume was less than 1,000 mL. If less than 1,000 mL, 

the volume was set equal to 1,000 mL.

Screening was done with one aliquant. A second aliquant was processed if the net reading of the first

aliquant was greater than or equal to a value in a chart that correlated with the volume of the 24-hr 

urine sample. A third aliquant was processed if the results (net readings) of the first two aliquants 

varied by 20% or more. The average result of the two aliquants that agreed was converted to 

μg/24-hr sample and reported only if the result was greater than or equal to the reporting level of 

5.8 μg/24-hr sample. Otherwise, the result was reported as background [100].

After April 1964, the urine sample was analyzed using the electroplating procedure described above 

for EU, and the results were reported in dpm/24-hr sample (or background) [101].

Gross Alpha (1952 to 1971)

Two methods were used to analyze urine samples for gross alpha counts from either plutonium or 

uranium. The ether extraction method was used from 1952 to December 12, 1962, and the tributyl 

phosphate (TBP) extraction method was used from December 12, 1962, to 1964. The TBP method 

was replaced by the TOPO method. All methods were nonspecific in extracting plutonium and 

uranium as well as americium and natural thorium [102].

In all methods, the entire urine sample was processed, and the final extract was evaporated on a 

planchet and counted on the gas-flow proportional counter. Counting time was typically 150 minutes, 

although from 1952 to 1955 count times of 55, 60, and 75 minutes, and in 1971 40 and 60 minutes, 

were also used.

Samples with results ≥0.9 dpm/24-hr sample were typically, but not always, counted using a PHA 

system to determine whether to credit the result to EU or to plutonium, or a portion to each. The 

default assumption through 1963 was to credit the result to EU unless the PHA count indicated 

otherwise. After 1963 (when EU operations were phased out), the default assumption was to credit 

the result to plutonium. In either case, the results should be considered upper bounds because of the 

nonspecificity of the analysis [103].

A.4 ASSESSMENT OF MDA

General Considerations

The MDA is assessed for plutonium, americium, EU, DU, and gross alpha, based on Equation A-1 

and the values of parameters for the methods. Some considerations are:

• The probabilities of Type I (false positive) and Type II errors (false negative) are each 5% 

(α = β = 0.05).

• The MDA is assessed for the typical, average, or median condition. If appropriate, the MDA is 

also assessed for the 5th- or 95th-percentile conditions.
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• The MDA is assessed for the methods as they should have been performed, with 

consideration of such factors as alpha transmission factor, blank subtraction, recovery fraction, 

and volume adjustment.

• For methods with two or more options in the same period (e.g., evaporation vs. 

electrodeposition, 40% detectors vs. 50% detectors), the option that gives the higher MDA is 

used.

The value of the MDA for the typical, average, or median condition pertains to the process and 

indicates the amount or activity in the population of urine samples that would have been detected with 

a 95% probability, given a properly set decision criterion that allows a 5% probability of a Type I error. 

In reality, the decision criterion (and method) at RFP was not based on the probability of a Type I 

error. Instead, an arbitrary level (10% of the tolerance level or any nonnegative value) was used as 

the decision criterion for recording and reporting detected amounts or activities.

The value of the MDA for the 5th- or 95th- percentile conditions pertains to individual samples for 

which the conditions of the sample (e.g., low volume) or conditions of the processing (low recovery, 

high blank, high alpha self-absorption) were marginal. The conditions of low recovery, low volume, 

and high alpha self-absorption are associated with the calibration factor K and can be incorporated 

either in the value of K or in the Δ value of ΔK.

Table A-4 lists sample volumes for routine 24-hour urine samples.

Table A-4. Sample volumes for

routine 24-hour urine samples.

Percentile Volume (mL)

5th 700

Median 1,350

95th 1,750

The values for the parameter values for the processes were obtained through review of the urine data 

logs for the periods from 1952 to 1955 and 1960 to 1971. For some years in these periods, logs for 

only a part of the year were available.

Data for Alpha-Counting Systems

Table A-5 lists the detector background (cpm) for the gas flow proportional counters, based on 

tabulations in the urine data logs from 1952 to 1955 and from 1960 to 1963, for a sample count time 

of 150 minutes [104]:

Table A-5. Detector background for gas flow proportional counters.

Date

Detector background (cpm)

Average 5th percentile Median 95th percentile

1950s 0.060±0.022 0.02 0.06 0.10

1960s 0.054±0.014 0.03 0.05 0.08

Composite 0.056±0.017 0.03 0.05 0.08

No documentation was found about the count time that was used to measure the detector 

background, but the count time was probably 150 minutes or longer. For the purpose of assessing 

the MDA, the composite average is used for the value of the detector background count rate BDet = 

0.056 cpm with the standard deviation sDet = 0.017 cpm for all alpha-counting methods (except for 
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americium) and for all sample count times [105]. For americium, the values for the 1960s are used 

because the americium process was not implemented in the 1950s [106].

The blank count rate was method-specific, and the application of the blank in the data analysis was 

variable between methods and within a method over time. A complication that was the intermittent, 

but persistent, was laboratory contamination artifacts that were introduced into blanks and worker 

samples. These artifacts caused false positives from a worker exposure viewpoint but real positives 

from a detection viewpoint. In practice, high blank values (a subjective decision) were generally 

ignored, and suspect (unexpectedly high) sample results were either confirmed or overruled by 

recounting, resampling, or analyzing another aliquant [107].

For the purpose of this MDA analysis, the median value of the blank is used to determine the process 

MDA and the 95th-percentile (low to high) value is used to determine the MDA for the more extreme 

conditions. Table A-6 summarizes the median and 95th-percentile blank count rates.

Table A-6. Median and 95th-percentile blank count rates.

Analyte Period

Blank count (cpm)

Median 95th percentile

Plutonium 1952–1971 0.06 0.28

EU 1952–1971 0.05 0.22

DU 1964–1971 0.05 0.22

Americium 1963–1971 0.07 0.26

Gross alpha 1952–1971 0.08 0.30

These values are the average of the yearly values extracted from available urine data logs (as 

reviewed by R. Falk in 2003 (the initial author of this TBD; see Section 5.7). For each of the analytes, 

the yearly median and 95th-percentile values did not differ enough over the period to warrant a 

separate MDA analysis. The blank values for EU and DU are based on log entries in 1963 and 1964 

for cell blank checks for the electrodeposition process.

The value of the blank count rate BBlk is taken from Table A-6 for the given analyte. The standard 

deviation sBlk is taken to be the square root of the blank count for the process divided by the count 

time of the process:

Blk

Blk

B T

s

T = (A-5)

The values for B, sB0, sB1, and so in the MDA equation (A-1) are derived from the detector background 

and blank values:

B TB B = + ( Det Blk ) (A-6)

2 2

B0 Det Blk s Ts s = + (A-7)

B1 s B = (A-8)

2 2

o B1 B0 s ss = + (A-9)
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For some analytes (EU, DU) and periods, the detection decision was based on the average of two 

aliquants. In this case, the value of so for the average of two aliquants is equal to the value of so for 

one aliquant divided by the square root of 2.

The value of ΔB is taken to be zero. This variable could be used to account for high blank values. 

Instead, the effect of high blank values is determined by using the 95th-percentile value of the blank.

The calibration factor K is a combination of the detector efficiency ε, the recovery R, and the volume 

adjustment factor (V/A). Also included is a factor that accounts for absorption of alpha particles in the 

residue of planchets or plates.

Common detectors were used for all alpha-counting methods. Table A-7 lists the efficiencies of the 

detectors (as noted above):

Table A-7. Efficiencies of alphacounting detectors.

Period Detector efficiency

1952–1953 0.45

1954–1963 0.50

1964–1971 0.40 and 0.50

For 1964 to 1971, the value of 0.40 is used as the efficiency for the MDA calculation [108].

Table A-8 lists the recoveries that were used in the MDA assessment, which are taken to be the 

median recovery and the 5th-percentile (low to high) value discerned from the spike data for the 

process.

Table A-8. Recoveries used in MDA assessments.

Analyte Period

Recovery

Median

5th 

percentile

Plutonium 1952–1962 0.57 0.25

Plutonium 1963–1971 0.67 0.28

EU 1952–1971 0.60 0.21

DU 1964–1971 0.60 0.21

Americium 1963–1965 0.67 0.29

Americium 1965–1971 0.80 0.26

Gross alpha 1952–1971 0.57 0.24

The recovery values are based on incomplete data sets and involve extrapolations to cover the total 

period. For plutonium from 1952 to 1962, the values are based on data for 1961 and 1962. For 

plutonium from 1963 to 1971, the values are based on data for 1963 to 1965 and 1969 to 1971. For 

EU and DU, recoveries were not calculated for the spiked samples. The median value is based on 

the value that was used for most of the period. The 5th-percentile value is based on the relative 

standard deviation (0.40) of the average count rate of the spiked samples from 1963 to 1966. For 

americium from 1963 to November 1, 1965, the values are based on a complete set for that period.

For 1965 to 1971, the values are based on data from November 1, 1965, to 1966, and 1968 to 1970. 

For gross alpha, the values are based on data from 1962 to 1969 for the TBP method. In general, 

values for all the processes are remarkably similar, except for americium from 1965 to 1971 [109].
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The volume adjustment factor V/A is incorporated into the calibration factor K as the reciprocal 

1/(V/A), so it becomes a multiplier with the efficiency and recovery. For convenience, the reciprocal of 

the volume adjustment factor is designated Vf.

For plutonium, americium, and gross alpha, the median condition is V = A and Vf = 1. The extreme 

condition is a low sample volume normalized to 1,200 mL: V = 1,200 mL, A = 700 mL (the 5thpercentile volume), and Vf = 0.58 [110].

For EU and DU (for the electrodeposition process), A = 50 mL, the median V = 1,350 mL, and Vf = 

0.037. The extreme condition is a high sample volume: V = 1,750 mL (the 95th-percentile volume), 

A = 50 mL, and Vf = 0.029 [111].

The absorption of the alpha particles in the residue that was evaporated on the planchets or 

electrodeposited on the plates should be incorporated into the value of the calibration factor. The 

factor to incorporate this effect is the fraction of the alphas that are emitted by the deposited analyte 

that successfully escape from the residue. Let this factor be designated Fa, where Fa = (1 – fraction of 

alphas absorbed in the residue), and let the fraction of alphas absorbed in the residue be fabs. 

Table A-9 lists the values of fabs, based on judgments of experienced bioassay chemists, for the 

extreme (95th-percentile) condition, and the corresponding values of Fa.

Table A-9. Fraction of alphas absorbed in residue.

Analyte Period

95th percentile

fabs Fa

Plutonium (evaporated) 1952–1962 0.4 0.6

Plutonium (evaporated) 1963–1971 0.3 0.7

Plutonium (electrodeposited) 1963–1971 0.05 0.95

EU (electrodeposited) 1952–1971 0.05 0.95

DU (electrodeposited) 1964–1971 0.05 0.95

Americium (evaporated) 1964–1971 0.3 0.7

Americium (electrodeposited) 1964–1971 0.05 0.95

Gross alpha (evaporated) 1952–1962 0.1 0.9

Gross alpha (evaporated) 1962–1971 0.3 0.7

From 1963 to 1971, approximately half of the plutonium and americium samples were 

electrodeposited. However, the identities of samples that were electrodeposited are not discernible 

from the databases and reports of urinalysis results that are readily accessible [112]. For the purpose 

of the MDA assessment, dose reconstructors should use the value of Fa for the evaporation process.

For the median condition, the value of Fa is taken to be 1 under the assumption that the absorption of 

alphas for the median condition of the planchet or plate was incorporated in the recovery value at the 

time.

The calibration factor K is the product of ε, R, Vf, and Fa:

K f a = εRV F (A-10)

The values of ΔB and ΔK are considered to be zero [113].
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Data for Fluorimetric Mass Measurements

Applying the MDA equation (A-1) to fluorimetric mass measurements involves setting the value of T to 

unity and eliminating the term “3”.

The value of sB0 is the standard deviation of the blank flux readings that are subtracted for the signal 

of the aliquant reading. The value of sB1 is set equal to sB0, and so is equal to the value of sB0 

multiplied by the square root of 2:

o B0 s s = 2 (A-11)

The value of sB0 was determined from a review by R. Falk of the urine data logs for 1955 and 1960 to 

1962. One discontinuity was noted on September 14, 1955. The value of sB0 before the discontinuity 

was 0.37 and, after the discontinuity, averaged 0.15.

The calibration factor K converts the fluorimeter net reading to the μg U/24-hr sample (see 

Equation A-4). In 1955, the calibration factor was applied to the uncorrected net reading. In 1960 and 

later, the calibration factor was applied to the corrected reading, which was the net reading multiplied 

by the factor 1.15 [114]. The factor 1.15 is incorporated into the value of K starting in 1960. For the 

1950s, the calibration factor for 1955 is used, as listed in Table A-10.

Table A-10. Gross alpha calibration factor.

Period K

1952–1959 75/V

1960–1964 87/V

For the median condition, the volume V is equal to 1,350 mL. For the extreme condition, the 95thpercentile volume of 1,750 mL is used.

The values of ΔB and ΔK are considered to be zero.

A.5 MDA VALUES

The value of the MDA is presented to two significant figures for information purposes. In most cases, 

the value of the MDA should be considered only to one significant figure.

Plutonium

The MDA for plutonium is assessed for the median condition and for the extreme (5th- or 95thpercentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for all assessments.

Table A-11 lists the values of the variables and the median MDA (dpm/24-hr sample).

Table A-12 lists the values of the variables for the extreme (5th- or 95th-percentile) conditions and the 

resulting MDA (dpm/24-hr sample) for all of the extreme conditions occurring for the same sample.

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. Table A-13 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.
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Table A-11. Values of variables and MDA for plutonium for median 

conditions.

Period

Values of the variables

MDA

(dpm/24-hr 

so ε R Vf Fa sample)

1952–1953 5.74 0.45 0.57 1.0 1.0 0.57

1954–1962 5.74 0.50 0.57 1.0 1.0 0.51

1963 5.74 0.50 0.67 1.0 1.0 0.44

1964–1971 5.74 0.40 0.67 1.0 1.0 0.54

Table A-12. Values of variables and MDA for plutonium for 

extreme conditions.

Period

Values of the variables

MDA

(dpm/24-hr 

so ε R Vf Fa sample)

1952–1953 7.98 0.45 0.25 0.58 0.6 5.0

1954–1962 7.98 0.50 0.25 0.58 0.6 4.5

1963 7.98 0.50 0.28 0.58 0.7 3.4

1964–1971 7.98 0.40 0.28 0.58 0.7 4.3

Table A-13. MDA for plutonium for one, two, or three 

extreme conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952–1953 0.76 1.3 0.98 0.95

1954–1962 0.68 1.2 0.88 0.85

1963 0.58 1.0 0.75 0.62

1964–1971 0.73 1.3 0.94 0.78

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952–1953 1.7 1.3 1.3 2.2 2.2 1.6

1954–1962 1.6 1.2 1.1 2.0 2.0 1.5

1963 1.4 1.0 0.97 1.8 1.5 1.1

1964–1971 1.7 1.3 1.2 2.3 1.9 1.3

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952–1953 3.0 2.9 2.2 3.7

1954–1962 2.7 2.6 2.0 3.4

1963 2.4 2.0 1.4 2.6

1964–1971 3.0 2.5 1.8 3.2

Enriched Uranium

The MDA for EU is assessed for the median condition and for the extreme (5th- or 95th-percentile) 

condition for the blank, the recovery, the volume factor Vf, and the alpha transmission factor Fa, 

individually and in combination. A count time of 150 minutes is used for MDA assessments from 1952 

to 1963.
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For 1964 to 1969, the count time of 30 minutes is used and, for 1970 to 1971, the count time of 

40 minutes is used. For 1952 to 1959, the value of so is calculated for one aliquant, and for 1960 to 

1971 the value of so is calculated based on the average of two aliquants.

Table A-14 lists the values of the variables and the median MDA (dpm/24-hr sample).

Table A-14. Values of variables and MDA for EU for median 

conditions.

Period

Values of the variables MDA

(dpm\24-hr 

so ε R Vf Fa sample)

1952–1953 5.45 0.45 0.60 0.037 1.0 14

1954–1959 5.45 0.50 0.60 0.037 1.0 13

1960–1963 3.85 0.50 0.60 0.037 1.0 9.4

1964–1969 1.57 0.40 0.60 0.037 1.0 31

1970–1971 1.83 0.40 0.60 0.037 1.0 25

Table A-15 lists the values of the variables for the extreme (5th- or 95th-percentile) conditions and the 

resulting MDA (dpm/24-hr sample) for all of the extreme conditions occurring for the same sample.

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. Table A-16 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-15. Values of variables and MDA for EU for extreme 

conditions.

Period

Values of the variables

MDA

(dpm\24-hr 

so ε R Vf Fa sample)

1952–1953 6.72 0.45 0.21 0.029 0.95 64

1954–1959 6.72 0.50 0.21 0.029 0.95 58

1960–1963 4.75 0.50 0.21 0.029 0.95 43

1964–1969 2.18 0.40 0.21 0.029 0.95 150

1970–1971 2.48 0.40 0.21 0.029 0.95 120

Depleted Uranium

The MDA for DU is assessed for two processes: fluorimetric mass measurements from 1952 to April 

30, 1964, and electrodeposition and alpha-counting measurements from May 1, 1964, to 1971.

For the fluorimetric mass measurements, the MDA is assessed for one aliquant because the decision 

for detection was based on one aliquant, even though quantification was based on the average of two 

aliquants [115]. In Table A-17, the MDA at the extreme condition is based on the 95th-percentile 

volume.

For the electrodeposition and alpha-counting measurements, the MDA values tabulated for EU for 

1964 to 1971 apply also to DU.
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Table A-16. MDA for EU for one, two, or three extreme 

conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952–1953 17 40 18 15

1954–1959 15 36 16 13

1960–1963 11 27 12 9.9

1964–1969 38 88 39 32

1970–1971 31 74 32 27

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952–1953 48 21 18 51 42 19

1954–1959 43 19 16 46 38 17

1960–1963 32 14 12 34 28 13

1964–1969 110 49 40 110 92 41

1970–1971 90 40 33 93 76 34

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952–1953 61 50 23 54

1954–1959 55 45 20 48

1960–1963 41 34 15 43

1964–1969 140 120 51 150

1970–1971 120 94 42 120

Table A-17. Values of variables and MDA for fluorimetric measurements 

of DU for median and extreme conditions.

Period

Values of the variables MDA (μg/24-hr sample)

sB0

Median 

K

Extreme 

K Median Extreme

1952–1955 0.37 0.056 0.043 31 40

1955–1959 0.15 0.056 0.043 12 16

1960–1964 0.15 0.064 0.050 11 14

Americium

The MDA for americium is assessed for the median condition and for the extreme (5th- or 95thpercentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for assessments from 

1963 to 1970. In 1971, the typical (and minimum) count time is 60 minutes.

Table A-18 lists the MDA to two significant figures.

Table A-19 lists the values of the variables for the extreme (5th- or 95th-percentile) conditions and the 

resulting MDA (dpm/24-hr sample) for all of the extreme conditions occurring for the same sample.
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Table A-18. Values of variables and MDA for americium for 

median conditions.

Period

Values of the variables MDA

(dpm/24-hr 

so ε R Vf Fa sample)

1963 5.82 0.50 0.67 1.0 1.0 0.44

1964–1965 5.82 0.40 0.67 1.0 1.0 0.55

1965–1970 5.82 0.40 0.80 1.0 1.0 0.46

1971 3.51 0.40 0.80 1.0 1.0 0.76

Table A-19. Values of variables and MDA for americium for 

extreme conditions.

Period

Values of the variables MDA

(dpm/24-hr 

so ε R Vf Fa sample)

1963 9.95 0.50 0.26 0.58 0.7 4.3

1964–1965 9.95 0.40 0.26 0.58 0.7 5.4

1965–1970 9.95 0.40 0.26 0.58 0.7 5.4

1971 5.94 0.40 0.26 0.58 0.7 8.9

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. Table A-20 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-20. Values of the MDA for americium for one, two, or 

three extreme conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1963 0.68 1.1 0.76 0.63

1964–1965 0.86 1.4 0.95 0.79

1965–1970 0.72 1.4 0.80 0.66

1971 1.2 2.3 1.3 1.1

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1963 1.8 1.2 0.98 2.0 1.6 1.1

1964–1965 2.2 1.5 1.2 2.4 2.0 1.4

1965–1970 2.2 1.2 1.0 2.4 2.0 1.1

1971 3.6 2.0 1.7 4.0 3.3 1.9

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1963 3.0 2.5 1.7 2.8

1964–1965 3.8 3.2 2.1 3.5

1965–1970 3.8 3.2 1.8 3.5

1971 6.2 5.2 2.9 5.7

Gross Alpha

The MDA for gross alpha measurements is assessed for the median condition and for the extreme 

(5th- or 95th-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha 
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transmission factor Fa, individually and in combination. A count time of 55 minutes is used for 1952,

75 minutes for 1953 to 1959, and 150 minutes for 1960 to 1971 for assessments of the MDA for both 

the median and extreme conditions, except for 1971, when a count time of 40 minutes is also used for 

the extreme condition. See Table A-21.

Table A-21. Values of variables and MDA for gross alpha 

measurements for median conditions.

Period

Values of the variables MDA

(dpm/24-hr 

ε R Vf Fa ε sample)

1952 3.26 0.45 0.57 1.0 1.0 1.0

1953 4.23 0.45 0.57 1.0 1.0 0.88

1954–1959 4.23 0.50 0.57 1.0 1.0 0.79

1960–1963 6.23 0.50 0.57 1.0 1.0 0.55

1964–1971 6.23 0.40 0.57 1.0 1.0 0.69

Table A-22 lists the values of the variables for the extreme (5th- or 95th-percentile) conditions and the 

resulting MDA (dpm/24-hr sample) for all of the extreme conditions occurring for the same sample.

Table A-22. Values of variables and MDA for gross alpha 

measurements for extreme conditions.

Period

Values of the variables MDA

(dpm\24-hr 

so ε R Vf Fa sample)

1952 6.09 0.45 0.24 0.58 0.9 7.4

1953 7.12 0.45 0.24 0.58 0.9 6.2

1954–1959 7.12 0.50 0.24 0.58 0.9 5.6

1960–1962 10.27 0.50 0.24 0.58 0.9 3.9

1963 10.27 0.50 0.24 0.58 0.7 5.0

1964–1971 10.27 0.40 0.24 0.58 0.7 6.3

1971 (T = 40 min) 5.18 0.40 0.24 0.58 0.7 13

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. Table A-23 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

A.6 DISCUSSION

The MDA is an a priori concept that can be applied a posteriori to a sample under certain 

circumstances: That the parameter values for the sample (e.g., volume, recovery, detector efficiency, 

count time) are or can be known before the processing of the sample result, and that the information 

is used conceptually to determine the subpopulation of conditions of which that sample is a member. 

Then the a priori MDA value for that subpopulation can be assigned to that sample. The sample 

volume, the characteristics of the detector that is used to count the sample, and the count time are all 
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Table A-23. Values of the MDA for gross alpha measurements for 

one, two, or three extreme conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952 1.6 2.5 1.8 1.2

1953 1.4 2.1 1.5 0.98

1954–1959 1.2 1.9 1.4 0.88

1960–1962 0.86 1.3 0.95 0.61

1963 0.86 1.3 0.95 0.79

1964–1971 1.1 1.6 1.2 0.98

1971 (T = 40 min) 2.2 3.4 2.4 2.0

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952 3.9 2.8 1.8 4.3 2.8 2.0

1953 3.3 2.4 1.5 3.6 2.3 1.7

1954–1959 2.9 2.1 1.4 3.2 2.1 1.5

1960–1962 2.0 1.5 0.96 2.3 1.5 1.1

1963 2.0 1.5 1.2 2.3 1.9 1.4

1964–1971 2.6 1.9 1.5 2.8 2.3 1.7

1971 (T = 40 min) 5.2 3.8 3.1 5.8 4.8 3.5

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952 6.7 4.3 3.1 4.7

1953 5.6 3.6 2.6 4.0

1954–1959 5.1 3.3 2.4 3.6

1960–1962 3.5 2.3 1.6 2.5

1963 3.5 2.9 2.1 3.1

1964–1971 4.4 3.6 2.6 4.0

1971 (T = 40 min) 9.0 7.5 5.4 8.3

known before the analysis of the sample measurement. In theory, but generally not in practice, the 

recovery could also be known before the analysis of the sample measurement.

The MDA values in this attachment represent overall process MDAs for the median and extreme 

conditions. However, sufficient information is presented to allow the determination of the MDA for a 

specific sample if the sample-specific parameter values are known. The sample-specific parameter 

values, except recovery, are generally recorded in the urine data logs, but not all of the urine data logs 

have been found and some might not have been archived.

The recoveries for 1952 to 1971 were determined by batch spikes. Not until 1973 were some 

plutonium samples spiked with an internal tracer (first 236Pu and later 242Pu). All plutonium samples 

were spiked with an internal tracer after 1978. Experience has shown that a significant variability of 

recovery can exist within a batch of samples. Therefore, the recovery of a batch spike does not 

necessarily indicate the recovery of each sample in the batch.

Whether to use the median or extreme value of the MDA or the extreme value depends on the 

purpose. By definition, the median value implies that half of the samples will have a sample-specific 

MDA that is lower, and half higher. If the purpose is to define a sample-specific conservative bound, 
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the MDA for the extreme condition should be considered. In general, the recovery fraction was the 

variable that had the most influence on the sample-specific MDA.
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TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Introduction ............................................................................................................................ 89

B.2 MDA Methodology.................................................................................................................. 89

B.3 History of Counting Systems and Procedures ........................................................................ 90

B.4 Assessment of MDA............................................................................................................... 93

B.5 Values of the Variables, 1969 – for NaI(Tl) and Phoswich Detector Systems......................... 95

B.6 Values of the Variables, 1976 – for Ortec Germanium Detector Systems............................... 96

B.7 Values of the Variables, 1978 – for PGT I Germanium Detector Systems .............................. 97

B.8 Values of the Variables, 1979 – for PGT II Germanium Detector Systems ............................. 99

B.9 Values of the Variables, 1985 – for PGT Organ Pipe Germanium Detector Systems ........... 100

B.10 Values of the Variables, 1985 – for EG&G Organ Pipe Germanium Detector Systems ........ 101

B.11 MDA for RFP Plutonium....................................................................................................... 102

LIST OF TABLES

TABLE TITLE PAGE

B-1 Discontinuity factors...............................................................................................................93

B-2 Calibration factors for the Ortec germanium detector system ................................................. 97

B-3 Calibration factors for the PGT I germanium detector system ................................................ 98

B-4 Calibration factors for the PGT II germanium detector system ............................................. 100

B-5 Values of variables for the PGT organ pipe germanium detector system ............................. 100

B-6 Calibration factors for the PGT organ pipe germanium detector system............................... 100

B-7 Values of variables for the EG&G organ pipe germanium detector system .......................... 101

B-8 Calibration factors for the EG&G organ pipe germanium detector system............................ 101

B-9 MDA conversion factors for values of ppm 241Am ................................................................. 102

B-10 Americium ingrowth in RFP plutonium.................................................................................. 103

B-11 Americium MDA for in vivo lung counts at RFP.................................................................... 104
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B.1 INTRODUCTION

In vivo lung counts have been performed at RFP since 1964 to determine the activity of plutonium in 

the lungs of workers who were exposed, or had the potential to be exposed, to airborne plutonium. 

The method of in vivo lung counts was to place one or more detectors over the chest of the subject 

and count the photons that were emitted from the plutonium mixture, if any, in the subject’s chest 

(Boss and Mann 1967). Plutonium was not detected directly because of the low abundance of 

gamma photons and because of the severe attenuation of the more abundant L X-rays. Instead, the 

59.5-keV gamma photon from 241Am was used as a surrogate. Americium-241 was present to some 

extent in all WG plutonium at RFP. The activity of plutonium was then calculated from the detected 

241Am by measuring, calculating, or assuming the fraction of the 241Am in the plutonium mixture on the 

date of the lung count. At RFP, the fraction of the 241Am in the plutonium mixture has historically been 

characterized in terms of parts per million by weight. Direct in vivo measurement of plutonium in the 

lungs, although investigated, was never implemented at RFP. The RFP lung counter detected 241Am. 

The assessment of the MDA, therefore, is focused on the MDA for 241Am. The MDA for plutonium can 

then be derived from the 241Am MDA based on the value of the ppm 241Am for the plutonium 

mixture. [116]

B.2 MDA METHODOLOGY

The general equation for the MDA is Equation 6 in the American National Standard, Performance 

Criteria for Radiobioassay (HPS 1996):

(1 2 23 Δ Δ B ks K B )( o ) MDA

KT

+ ++ = (B-1)

where:

K = calibration factor

ΔK = the maximum fractional systematic error bound in the calibration factor K

B = the total count of the appropriate blank

ΔB = the maximum expected fractional systematic error bound in the appropriate blank

k = the abscissa of the standardized normal distribution corresponding to the 0.05 

probability level (for α = 0.05 and β = 0.05, k = 1.645)

so = the standard deviation in the net sample count of a subject with no additional analyte

T = the standard subject counting time for the procedure:

2

2 B0

o B1 2

s

s s

m

= + (B-2)

where:

sB1 = the standard deviation of the subject, where the subject contains no actual 

analyte above that of the appropriate blank

sB0 = the standard deviation in the unadjusted count of the appropriate blank

m = the adjustment factor for the appropriate blank

Applying this equation to in vivo lung counting at RFP involves determining the value of each of these 

variables for the counting systems and procedures used at RFP as the systems and procedures 
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evolved. The MDA for in vivo measurements is necessarily individual-specific because the 

detectability of 241Am in the chest is a significant function of the CWT of the subject.

The MDA can also be determined empirically from replicate measurements on an appropriate blank. 

This approach is used for the systems starting in 1995 at RFP.

B.3 HISTORY OF COUNTING SYSTEMS AND PROCEDURES

The in vivo lung-counting systems at RFP consisted of photon detectors in a shielded room 

(6-in.-thick low-background steel lined with layers of lead, tin, and zinc) with electronic equipment 

(amplifiers and multichannel analyzers) to process and record the data.

There were three counting rooms:

• Room A, built in 1964, operational in 1965;

• Room B, built in 1968, operational in 1969; and

• Room C, built in 1975, operational in 1976 [117]

Each room was equipped with a detector system. When a new detector system was implemented, 

the previous system was usually maintained as a backup system. As a result, end dates for use of a 

given detector system are not known. In the era of the germanium detector systems, two or more 

detector systems could have been operational simultaneously. In that situation, the detector system 

is identified in the record for each lung count.

1964 to 1968 [118]

There was one counting room. The detector system consisted of two NaI(Tl) scintillation detectors 

(there was a third detector used for cesium and potassium measurements); each detector was round 

with a diameter of 4 in. and was 4 mm thick with a surface area of 80 cm2

. These detectors were 

known as the “4×4 detectors.” In most situations, the detectors were configured with one detector 

above the left portion of the upper chest; the second detector was over the liver and gut region. The 

chest detector was sometimes placed over the right portion of the upper chest rather than the left 

position. In other cases, both detectors were placed over the chest. The chest detectors were placed 

in a framework called a jig to allow a standard and reproducible position for all subjects. Count time 

was either 40 MLT or 20 MLT. Two backgrounds were used: (1) room background and (2) matched 

subject background. The room background was the count rate in the empty counting room at the start 

of the day. The matched subject background was the count rate of an unexposed subject with 

matched 137Cs and 40K count rates. Calibration was based on 241Am-impregnated epoxy lungs in the 

chest cavity of a water-filled REMAB phantom from Alderson Research Laboratories. No adjustment 

was made for CWT.

1969 to 1976 [119]

During this period, two counting rooms were operational with three 4- by 4-in. NaI(Tl) scintillation 

detectors, two over the upper chest (right and left portions) and one over the liver/gut region. The 

liver/gut detector was eliminated in 1974.

Changes:

1. The ROI of the 59.5-keV photopeak of 241Am was expanded.
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2. The use of the jig for positioning the detectors was discontinued. Instead, the detectors were 

positioned in light contact with the surface of the chest.

3. The standard count time was changed to 2,000 seconds (1,000 seconds for expedited counts).

4. The method of the matched subject background based on 137Cs and 40K was replaced by the 

index method.

The index method had the following features (Bistline 1968):

1. Subjects were characterized by an index I equal to the ratio of the subject’s weight (W, 

pounds) divided by twice the subject’s height (H, inches).

2. A population of at least 20 known cold (unexposed) subjects of a diversity of indices was 

counted to generate a data set of net count rate versus index.

3. A curve fit to the data set generated a prediction equation with the index as the variable.

4. The subject’s index was used to determine the predicted net count rate for the subject.

This approach was applied separately for the right chest, the left chest, and the liver/gut.

In 1973, a phoswich detector system [a detector with a primary scintillation Nal(TI) layer backed by a 

Csl layer for coincidence counting] was implemented and used intermittently into the 1980s. The 

NaI(Tl) layer of the phoswich detectors was dimensionally the same as the 4- by 4-in. detectors.

This system lacked the stability of the NaI(Tl) detector system and was used mainly as a backup 

system. Use of the phoswich system to detect plutonium directly using the plutonium L X-rays was 

not successfully implemented at RFP.

In about 1972, room background was measured at the start of the day shift, at noon, and at the start 

of the night shift. The value of the room background RFP used was the five-point moving average of 

the last five counts.

Starting in 1969 [120], the ppm 241Am was measured routinely from a representative sample of the 

plutonium mixture associated with incidents with the potential for inhalation exposure of workers. This 

situation was called a “PI” (for possible inhalation) and refers both to the incident and to the worker 

involved in the incident.

In this period, the use of a lithium-drifted germanium detector system was investigated but was never 

implemented.

1976 to 1985 [121]

This period is the era of the high-purity germanium detector array systems. Three counting rooms 

were operational. When the germanium systems were implemented, most, if not all, quantitative 

measures were accomplished with that system. The NaI(Tl) and phoswich systems were used only 

as screening systems, and later only as backup systems. The germanium systems in this period 

featured four detectors mounted in an array attached to a single cryostat containing liquid nitrogen. 

The system had two of these arrays, one over the upper right chest and the other over the upper left 

chest. A full system consisted of eight detectors. However, occasionally one or more of the detectors 

failed and were electronically eliminated from the system. A minimum system was five detectors, 
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three in the right array and two in the left. To maintain a minimum functional system, a hybrid system 

consisting of two arrays of different characteristics was frequently used.

The germanium system implementation timeline was:

1976 Ortec detectors, 10 cm2 per detector, two arrays,

1977 PGT I detectors, 15 cm2 per detector, two arrays,

1979 First array, PGT II detectors, 18 cm2 per detector, and

1980 Second array, PGT II detectors, 18 cm2 per detector.

Other changes in this period were:

1. The calibration factor for the germanium systems was adjusted for the CWT of the subject. 

The thickness (centimeters) was equal to twice the index value minus 0.1 (CWT = 2I – 0.1).

2. Calibration was accomplished using a Masonite phantom from 1976 to 1978.

3. Calibration was accomplished using the LLNL phantom starting in 1979.

4. The method of determining the background changed for the germanium systems. Room and 

subject background were determined as a unit from the subject’s own spectrum using an ROI 

in the range of 65 to 72 keV.

1985 to 1995 [122]

In this period, germanium detectors in an organ pipe configuration were implemented. Instead of 

clustering four detectors in an array with a common cryostat, each detector was attached to its own 

cryostat, which was tall and slender. The detectors with the cryostats were then clustered in arrays, 

two to four detectors per array, over the right and left portions of the upper chest. If a detector 

malfunctioned, it was physically replaced with a backup functional detector. A minimum system from 

1985 to 1991 was five detectors, three on the right and two on the left. The full system was seven 

detectors, four on the right and three on the left, although the routine system generally consisted of 

six, either four on the right and two on the left or three on each side. In 1991, the full system was six 

detectors with either four on the right and two on the left or three on each side.

The germanium system implementation timeline was:

1985 PGT organ pipe detectors, 20 cm2 per detector

1991 EG&G Ortec organ pipe detectors, 20 cm2 per detector

No other significant changes were made during this period.

1995 to 2005 [123]

In May 1995, the lung counter hardware, software, and detectors were upgraded. The data 

acquisition and analysis were accomplished using the Canberra Industries program ABACOS-Plus. 

Instead of the ROI approach that was used previously, this program used a peak-search method to 

detect activity of a radionuclide. The value of the MDA was established by replicate measurements 

on an appropriate blank. The germanium detectors were replaced by EG&G Ortec organ pipe 

detectors with 38 cm2 per detector. The standard system was four detectors, two on each side. The 

minimum system was three detectors, two on the right and one on the left.
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Another significant change (RFETS 2000b) was the equation to determine CWT. ABACOS-Plus

incorporates the equation developed at LLNL:

1 973 (cm) 2 0038 . W CWT . H = − (B-3)

where:

W = subject’s weight (pounds)

H = subject’s height (inches)

The effect of this change is an adjustment factor given by:

0 635 0 5364 . I CWT Adjustment Factor . e = (B-4)

This adjustment factor is a multiplier to the activity of 241Am, detected using the 59.5-keV gamma, for 

all previous detector systems at RFP. Equation B-4 can also be applied as a divisor to calibration 

factors for previous systems at RFP.

B.4 ASSESSMENT OF MDA

The value of the MDA for 241Am is assessed here for each detector system and for each significant 

change in the procedure. It is assessed not only for the typical RFP male (I = 1.35, CWT = 3.3 cm) 

but also for a reasonable range of statures (I = 0.90, CWT = 1.5 cm and I = 1.80, CWT = 5.1 cm). 

The assessment is also done for the minimally configured system as well as for the standard system 

and for half of the normal count time (for expedited lung counts) as well as the full count time.

Discontinuities, which were significant changes in methods that affected the interpretation of the raw 

data (and therefore the MDA), were identified through review of available records and were 

incorporated into the value of the calibration factor. This process was done starting with the most 

recent calibration method, assumed to be the most accurate. The factors for each discontinuity were 

then applied as divisors to the calibration factor through the history of the systems. As an alternative, 

the product of the factors, for the appropriate period, can be used in place of the term (1 + ΔK) in 

Equation B-1. Table B-1 lists the discontinuity factors.

Table B-1. Discontinuity factors.

Year Discontinuity Factor

1995 New CWT method

Index = 0.90 0.95

Index = 1.35 1.26

Index = 1.80 1.68

1979 Calibration using LLNL phantom 1.30

1969 Fixed positioning discontinued

ROI for 59.5-keV photopeak increased

1.45

The discontinuity factors for the CWT can be calculated by any index I using:

0 635 0 5364 . I CWT Discontinuity Factor . e = (B-5)
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Values of the Variables, 1964–1968

The minimum system was one NaI(Tl) detector over the left chest.

Count time T = 20 MLT or 40 MLT [124]

The appropriate blank B was the net subject background (after room background was subtracted) 

estimated from matched unexposed subjects based on 137Cs and 40K measurements.

B = 600 for T = 20 MLT

B = 1,200 for T = 40 MLT

ΔB = 0.2, estimated as the upper bound for this method

The value of so is calculated from counting statistics, including the total subject count, which is taken 

as the sum of B and the room background R.

R = 500 for T = 20 MLT

R = 1,000 for T = 40 MLT

Because the decision of detection was based on the comparison of the net subject count rate (after 

subtraction of room background) with the predicted net count rate of the appropriate blank, the 

calculation of sB1 includes an extra component of the room background.

sB1

2 = Total subject count + R = B + 2R

= 1,600 for T = 20 MLT

= 3,200 for T = 40 MLT

sB0

2 = B

= 600 for T = 20 MLT

= 1,200 for T = 40 MLT

m = 1

so = 44.9 for T = 20 MLT

= 66.3 for T = 40 MLT

The 241Am calibration factor K for two detectors, normalized to the calibration with the LLNL phantom 

[125] and incorporating the discontinuity factors (Equation B-5, 1.30, and 1.45) is given by:

0 2359 2 0 1 ( )

0 635

55 13 . I.

. I

. e K

e

− −

= (B-6)

The calibration factor for the system with only one detector over the left portion of the chest is given 

by Equation B-6 multiplied by 0.43. This factor is the fraction of the total activity in the calibration 

lungs of the RFP LLNL phantom that is in the left portion of the lung. The MDA, therefore, pertains to 

the activity in the total lung based on the detection of activity only in the left portion of the lung.

K = 8.96 for I = 0.90

K = 5.45 for I = 1.35

K = 3.31 for I = 1.80

Because K is normalized to the calibration with the LLNL phantom and the discontinuity factors are 

incorporated into K, the value of ΔK is taken to be zero. Because the term (1 + ΔK) in Equation B-1 is 

a multiplier to the MDA and because the value of ΔK is estimated based on the professional judgment 
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of the analyst, one can easily adjust the values of the MDA in this attachment if another analyst has a 

different judgment.

For the standard system of two detectors, over both the right and left portions of the lungs, the counts 

are basically doubled and the values of the variables are:

B = 1,200 for T = 20 MLT

B = 2,400 for T = 40 MLT

ΔB = 0.2, estimated as the upper bound for this method

R = 1,000 for T = 20 MLT

R = 2,000 for T = 40 MLT

sB1

2 = Total subject count + R = B + 2R

= 3,200 for T = 20 MLT

= 6,400 for T = 40 MLT

sB0

2 = B

= 1,200 for T = 20 MLT

= 2,400 for T = 40 MLT

m = 1

so = 69.3 for T = 20 MLT

= 93.8 for T = 40 MLT

K = 20.85 for I = 0.90

= 12.67 for I = 1.35

= 7.70 for I = 1.80

B.5 VALUES OF THE VARIABLES, 1969 – FOR NAI(TL) AND PHOSWICH DETECTOR 

SYSTEMS

The standard system was two detectors over the left and right portions of the chest. This is also the 

minimum system.

Count time T = 1,000 seconds or 2,000 seconds [126]

The appropriate blank was the net subject background (after room background was subtracted) 

estimated from matched, unexposed subjects based on the subject’s index:

B = 1,100 for T = 1,000 seconds

B = 2,200 for T = 2,000 seconds

ΔB = 0 for the NaI(Tl) detector system

ΔB = 0.1, estimated for the phoswich detector system, because the system was less stable 

than the NaI(Tl) detector system

The value of so is calculated from counting statistics, including the total subject count, which is taken 

as the sum of B and the room background R. The value of sB0 is taken to be 10% of the value B, 

based on the typical relative standard deviation of the predicted subject net count rate.

R = 833 for T = 1,000 seconds

R = 1,667 for T = 2,000 seconds

sB1

2 = Total subject count + R = B + 2R

= 2,767 for T = 1,000 seconds

= 5,533 for T = 2,000 seconds
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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

sB0

2 = (0.1B)

2

= 12,100 for T = 1,000 seconds

= 48,400 for T = 2,000 seconds

m = 1

so = 121.9 for T = 1,000 seconds

= 232.2 for T = 2,000 seconds

The 241Am calibration factor K for two detectors, normalized to the calibration with the LLNL phantom 

and incorporating the discontinuity factors (Equation B-5 and 1.30) is given by:

0 2359 2 0 1 ( )

0 635

79 94 . I.

. I

. e K

e

− −

= (B-7)

Which yields the following results:

K = 30.23 for I = 0.90

= 18.37 for I = 1.35

= 11.16 for I = 1.80

B.6 VALUES OF THE VARIABLES, 1976 – FOR ORTEC GERMANIUM DETECTOR SYSTEMS

The standard system was two arrays, each array with four detectors over the left and right portions of 

the chest. The minimum system was two arrays with a total of eight detectors.

Count time T = 1,000 seconds or 2,000 seconds

The appropriate blank was the count in the subject’s spectrum (composite for all detectors) in the 

range of 65 keV to 72 keV, divided by eight. The subject, in essence, was his own blank with 

essentially no bias. Room background was no longer assessed separately for germanium systems.

ΔB = 0

m = 8

For eight detectors:

B = 341 for T = 1,000 seconds (unadjusted by m = 8)

B = 682 for T = 2,000 seconds (unadjusted by m = 8)

For five detectors:

B = 213 for T = 1,000 seconds (unadjusted by m = 8)

B = 427 for T = 2,000 seconds (unadjusted by m = 8)

For the calculation of sB1, the subject background is B/8.

For eight detectors:

T = 1,000 seconds:

sB1 = 6.53 sB0 = 18.5 so = 6.93
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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

T = 2,000 seconds:

sB1 = 9.23 sB0 = 26.1 so = 9.79

For five detectors:

T = 1,000 seconds:

sB1 = 5.17 sB0 = 14.6 so = 5.48

T = 2,000 seconds:

sB1 = 7.30 sB0 = 20.7 so = 7.75

The 241Am calibration factor K for two arrays with a total of eight detectors, based on the calibration 

with the LLNL phantom and incorporating the discontinuity factors [Equation B-5 and 1.30 (for pre1979 systems)] is given by:

0 3398 2 0 1 ( )

0 635

24 12 . I.

. I

. e K

e

− −

= (B-8)

and, for Ortec systems 1979 and after:

0 3398 2 0 1 ( )

0 635

31 36 . I.

. I

. e K

e

− −

= (B-9)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

Table B-2 lists the calibration factors for the Ortec germanium detector system.

Table B-2. Calibration factors for the Ortec 

germanium detector system.

Index

Eight-detector calibration factor (K)

Pre-1979 1979

0.90 7.64 9.94

1.35 4.23 5.50

1.80 2.34 3.04

B.7 VALUES OF THE VARIABLES, 1978 – FOR PGT I GERMANIUM DETECTOR SYSTEMS

The PGT I germanium system is basically the same as the Ortec germanium system. The main 

difference is that the PGT I detectors had a larger surface area but a poorer resolution [127].

ΔB = 0

m = 4

For eight detectors:

B = 240 for T = 1,000 seconds (unadjusted by m = 4)

B = 480 for T = 2,000 seconds (unadjusted by m = 4)
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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

For five detectors:

B = 150 for T = 1,000 seconds (unadjusted by m = 4)

B = 300 for T = 2,000 seconds (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 seconds:

sB1 = 7.75 sB0 = 15.5 so = 8.67

T = 2,000 seconds:

sB1 = 10.95 sB0 = 21.9 so = 12.2

For five detectors:

T = 1,000 seconds:

sB1 = 6.12 sB0 = 12.2 so = 6.84

T = 2,000 seconds:

sB1 = 8.66 sB0 = 17.3 so = 9.68

The 241Am calibration factor K for two arrays with a total of eight detectors, based on the calibration 

with the LLNL phantom and incorporating the discontinuity factors [Equation B-5 and 1.30 (for pre1979 systems)] is given by:

0 3292 2 0 1 ( )

0 635

34 09 . I.

. I

. e K

e

− −

= (B-10)

and, for PGT I systems 1979 and after:

0 3292 2 0 1 ( )

0 635

44 318 . I.

. I

. e K

e

− −

= (B-11)

For the minimum system of five detectors, adjust the calibration factor by multiplying by 0.625 (5/8).

Table B-3 lists calibration factors for the PGT I germanium detector system.

Table B-3. Calibration factors for the PGT I 

germanium detector system.

Index

Eight-detector calibration factor (K)

Pre-1979 1979 →

0.90 11.00 14.30

1.35 6.15 7.99

1.80 3.43 4.46
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ATTACHMENT B

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

B.8 VALUES OF THE VARIABLES, 1979 – FOR PGT II GERMANIUM DETECTOR SYSTEMS

The PGT II germanium system is basically the same as the Ortec and PGT I systems. The main 

difference is that the PGT II detectors, again, had a larger surface area but a poorer resolution [128].

ΔB = 0

m = 4

For eight detectors:

B = 273 for T = 1,000 seconds (unadjusted by m = 4)

B = 546 for T = 2,000 seconds (unadjusted by m = 4)

For five detectors:

B = 170 for T = 1,000 seconds (unadjusted by m = 4)

B = 341 for T = 2,000 seconds (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 seconds:

sB1 = 8.26 sB0 = 16.5 so = 9.23

T = 2,000 seconds:

sB1 = 11.7 sB0 = 23.4 so = 13.1

For five detectors:

T = 1,000 seconds:

sB1 = 6.53 sB0 = 13.1 so = 7.31

T = 2,000 seconds:

sB1 = 9.23 sB0 = 18.5 so = 10.3

The 241Am calibration factor K for two arrays with a total of eight detectors (incorporating 

Equation B-5), is given by:

0 3579 2 0 1 ( )

0 635

38 65 . I.

. I

. e K

e

− −

= (B-12)

For the minimum system of five detectors, adjust the calibration factor by multiplying by 0.625 (5/8).

Table B-4 lists calibration factors for the PGT II germanium detector system.
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ATTACHMENT B

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

Table B-4. Calibration factors for the PGT II 

germanium detector system.

Index

Eight-detector calibration factor 

(K)

0.90 11.88

1.35 6.47

1.80 3.52

B.9 VALUES OF THE VARIABLES, 1985 – FOR PGT ORGAN PIPE GERMANIUM DETECTOR 

SYSTEMS

The PGT organ pipe germanium system is basically the same as the previous germanium array 

systems. The main difference is the ability to maintain a stable, standard configuration with six 

detectors.

ΔB = 0

m = 4

Table B-5 lists the values of variables for the PGT organ pipe germanium detector system.

Table B-5. Values of variables for the PGT 

organ pipe germanium detector system 

T = 1,000 s T = 2,000 s

B 215 429

sB1 7.33 10.4

sB0 14.7 20.7

so 8.20 11.6

The 241Am calibration factor K for two arrays with a total of six detectors (incorporating Equation B-5) 

is given by:

0 2946 2 0 1 ( )

0 635

34 32 . I.

. I

. e K

e

− −

= (B-13)

Table B-6 lists calibration factors for the PGT organ pipe germanium detector system.

Table B-6. Calibration factors for the PGT 

organ pipe germanium detector system.

Index

Six-detector calibration factor 

(K)

0.90 11.74

1.35 6.77

1.80 3.90
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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

B.10 VALUES OF THE VARIABLES, 1985 – FOR EG&G ORGAN PIPE GERMANIUM 

DETECTOR SYSTEMS

The EG&G organ pipe germanium system is basically the same as the previous PGT organ pipe 

germanium array system.

ΔB = 0

m = 4

Table B-7 lists the values of variables for the EG&G organ pipe germanium detector system.

Table B-7. Values of variables for the EG&G 

organ pipe germanium detector system.

T = 1,000 s T = 2,000 s

B 204 408

sB1 7.14 10.1

sB0 14.3 20.2

so 7.98 11.3

The 241Am calibration factor K for two arrays with a total of six detectors, incorporating Equation B-5, 

is given by:

0 3708 2 0 1 ( )

0 635

42 36 . I.

. I

. e K

e

− −

= (B-14)

Table B-8 lists calibration factors for the EG&G organ pipe GE detector system.

Table B-8. Calibration factors for the EG&G 

organ pipe germanium detector system.

Index

Six-detector calibration factor 

(K)

0.90 12.73

1.35 6.85

1.80 3.69

Values of the Variables, 1995

The MDA for the system at RFP was not determined analytically using Equation B-1. Instead, the 

MDA was determined empirically from replicate measurements on an appropriate blank that simulated 

the counts of the average RFP worker (CWT = 3.36 cm). Therefore, there are no values of the 

variables to be listed here. The value of the MDA for the average RFP worker (CWT = 3.36 cm, I = 

1.35) is 0.3 nCi 241Am.

To extrapolate this value to the range of workers (CWT = 1.15 cm, I = 0.90 to CWT = 5.10 cm, 

I = 1.80), the following approach was used to establish the calibration factor equation as a function of 

CWT. The efficiency equation is:

2

1

a CWT ε a = (B-15)
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ATTACHMENT B

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

where

ε = counts per minute per gamma from 241Am

a1 = 0.045 (factor determined from calibration)

a2 = -0.41 (factor determined from calibration)

The efficiency equation converts to the style of historical calibration equations using the conversion 

factors of 0.359 gamma photons (59.5 keV) per 241Am nuclear transformation and 797 γ/min per nCi 

241Am. The derived calibration equation is:

0 41 35 9 . CWT K . − = (B-16)

The MDA for any value of CWT is then obtained from the product of 0.3 nCi (the MDA for the average 

RFP worker) and the ratio (9.05/K for the value of CWT).

B.11 MDA FOR RFP PLUTONIUM

The MDA for RFP plutonium is derived from the MDA of 241Am based on the value of the ppm 241Am in 

the plutonium mixture at the time of the lung count. To convert the MDA for 241Am to the MDA for 

plutonium (239Pu and 240Pu), the MDA for 241Am is multiplied by the factor:

6 241

 241

1 10 Am 48 2 Am

ppm MDA Conversion Factor

. ppm

× − = (B-17)

Table B-9 lists MDA conversion factors for some typical values of ppm 241Am.

Table B-9. MDA conversion factors for 

values of ppm 241Am.

ppm Am-241

MDA conversion 

factors

100 207

1,000 20.7

10,000 2.05

The task is to determine the value of the ppm 241Am at the time of the lung count. The practice at 

RFP was to measure the ppm 241Am in a representative sample of material from a possible inhalation 

incident. If a representative sample was not obtained or the origin of the intake was not known, a 

default value of 1,000 ppm 241Am was used and was assigned to the date of the intake or to the date 

of the first positive lung count if the date of the intake was not known. For subsequent lung counts, 

the value of the ppm 241Am was updated to account for the ingrowth of the 241Am from the nuclear 

transformation of 241Pu and for the radioactive decay of the 241Am. The rate of ingrowth of 241Am in 

the plutonium mixture depends on the fraction by weight of the 241Pu in the mixture. The initial weight 

fraction of 241Pu in RFP plutonium was taken to be 0.005 in the 1950s and 1960s and 0.0036 in the 

1970s and later [129]. Table B-10 lists values of the ppm 241Am at times (years) after the intake for 

initial values of ppm 241Am of 100, 1,000, and 10,000. The value of 100 ppm 241Am can be taken as 

the lower bound [130] and represents freshly purified plutonium.
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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

Table B-10. Americium ingrowth in RFP plutonium.

Years

Am-241 ppm at time (yr) after intake

Initial fraction Pu-241 = 0.0036 Initial fraction Pu-241 = 0.0050

100 1,000 10,000 100 1,000 10,000

1 270 1,200 10,200 340 1,200 10,200

2 430 1,300 10,300 560 1,500 10,400

4 730 1,600 10,600 980 1,900 10,800

6 1,000 1,900 10,800 1,400 2,200 11,100

10 1,500 2,400 11,200 2,000 2,900 11,700

20 2,300 3,200 11,900 3,100 4,000 12,700

30 2,800 3,600 12,200 3,800 4,700 13,200

40 3,000 3,900 12,300 4,200 5,000 13,500

50 3,200 4,000 12,300 4,400 5,200 13,500

The appropriate value of the ppm 241Am should be applied for lung counts that occurred after a known 

or assumed intake.

The equation to calculate the ppm 241Am for any time (years) after the intake is:

( ) Pu241 Am241

Pu239

6

0 2

1 0

0 2

λ T λ T 1 10

λ T

A L A LP e e

AL e

− −

−

×

= −+

+ (B-18)

where

A = ppm 241Am at time T (years)

L1 = λPu241 ÷ (λAm241 – λPu241)

P0 = initial 241Pu ppm = (initial 241Pu fraction by weight) × (1 × 106 – A0)

λPu241 = decay constant for 241Pu (half-life = 14.4 years) = 0.0481

λAm241 = decay constant for 241Am (half-life = 433 years) =0.00160

A0 = initial ppm 241Am

L2 = Am241

6 1 10 0

λ T e

A

−

− × −

(B-19)

λPu239 = decay constant for 239Pu (half-life = 24,100 years) = 0.0000288

Half-times are from Table of Isotopes, Seventh Edition (Lederer and Shirley 1978).

Table B-11 summarizes the americium MDAs for RFP in vivo lung counts.
ATTACHMENT B

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 104 of 177

Table B-11. Americium MDA for in vivo lung counts at RFP.a

Period

Detector 

system Indexb

MDA (nCi) for Am-241c,d

Comments

Minimum system Standard system

Half timee Full time Half timee Full time

1964–

1968

NaI(Tl) 4×4 0.90 1.7 1.5 1.3 1.2 Full time = 40 MLT.

Minimum system is one detector over the left chest.

Standard system is two detectors, over right and left chests.

1.35 2.8 2.5 2.1 1.9

1.80 4.6 4.1 3.5 3.2

1969– 0.90 – – 0.80 0.76 Full time = 2,000 s.

1.35 – – 1.3 1.3 Standard system is two detectors, over right and left chests.

1.80 – – 2.2 2.0

1973– Phoswich 0.90 – – 1.2 1.2 Full time = 2,000 s.

Standard system is two detectors, over right and left chests.

NaI sensitive layer is the same as the NaI 4×4.

1.35 – – 2.0 2.0

1.80 – – 3.3 3.2

1976–

1978f

Ortec Arrays 

(High-purity 

Ge)

0.90 0.26 0.18 0.20 0.14 Full time = 2,000 s.

Standard system is eight detectors in two arrays.

Minimum system is five detectors in two arrays.

1.35 0.48 0.32 0.37 0.25

1.80 0.86 0.59 0.66 0.45

1979– 0.90 0.20 0.14 0.16 0.11

1.35 0.37 0.25 0.28 0.19

1.80 0.66 0.45 0.51 0.35

1978– PGT I Arrays 

(High-purity 

Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.40 0.27 0.31 0.21

1.80 0.71 0.49 0.55 0.38

1979– 0.90 0.17 0.12 0.13 0.09

1.35 0.31 0.21 0.24 0.16

1.80 0.55 0.38 0.42 0.29

1979– PGT II Arrays

(High-purity 

Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.40 0.28 0.31 0.21

1.80 0.74 0.50 0.57 0.39

1985– PGT Organ 

Pipe (OP) Ge 

Detectors

0.90 – – 0.15 0.11 Standard system = six detectors.

Standard count time = 2,000 s.

Occasionally, five or seven detectors were used.

1.35 – – 0.26 0.18

1.80 – – 0.46 0.32

1991– EG&G Organ 

Pipe Ge 

Detectors

0.90 – – 0.14 0.10 Standard system = six detectors.

1.35 – – 0.26 0.18 Standard count time = 2,000 s.

1.80 – – 0.48 0.33

1995– Ortec 2 

Organ Pipe 

Ge Detectors

0.90 – – – 0.14 Standard system = four detectors.

1.35 – – – 0.3 Standard count time = 2,000 s.

1.80 – – – 0.6

a. Americium-241 grows into the plutonium mixture from the nuclear transformation of 241Pu. The initial weight fraction of 241Pu in RFP plutonium was 0.0050 in the 1950s and 1960s and 

0.0036 in the 1970s and 1980s. For freshly purified plutonium, with a residual of approximately 100 ppm 241Am, the ppm 241Am would be 270 to 340 after the 1st year, 430 to 560 after the 
ATTACHMENT B

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (continued)

Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 105 of 177

2nd year, 730 to 980 after the 4th year, 1,000 to 1,400 after the 6th year, 1,500 to 2,000 after the 10th year, 2,300 to 3,100 after the 20th year, 2,800 to 3,800 after the 30th year, 3,000 to 

4,200 after the 40th year, and 3,200 to 4,400 after the 50th year.

b. The index is the ratio of the weight (pounds) of the subject divided by twice the height (inches) and is correlated with the CWT. The index of 1.35 represents the typical RFP male subject, 

with a reasonable range of 0.90 (CWT = 1.5 cm) to 1.80 (CWT = 5.1 cm).

c. To convert the MDA for 241Am to the MDA for 239,240Pu, multiply the MDA for 241Am by [(1 × 106 – ppm 241Am) ÷ (48.2 × ppm 241Am)], where ppm 241Am is the parts per million by weight of 

the 241Am in the plutonium mixture at the time of the lung count.

d. – = not applicable.

e. Halved count times were usually used for nonscheduled counts or when a large number of subjects needed to be counted expeditiously.

f. Starting in 1978, hybrid germanium systems were used that combined two different germanium arrays or detector types. For hybrid systems, use the higher of the MDA values for the 

involved detector types.
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ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Urinalysis Record Card and HSDS – Urinalysis Detail (1) (first activity date on the HSDS 

portion: 10-29-54)................................................................................................................ 107

C-2 Urinalysis Record Card and HSDS – Urinalysis Detail (2) (first activity date on the HSDS 

portion 8-19-53) ................................................................................................................... 108

C-3 Urinalysis Record Card and HSDS – Urinalysis Detail (3) (first activity date on the HSDS 

portion: 1-6-58) ................................................................................................................... 109

C-4 HSDS – Urinalysis Detail (1) (first activity date 9-17-58) ...................................................... 110

C-5 HSDS – Urinalysis Detail (2) (first activity date 3-19-73) ...................................................... 111

C-6 Analytical Report – Bioassay Analysis Data 3-15-93............................................................ 112

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 .......................................................... 113

C-8 Form 1 – Sample Results 1-29-96........................................................................................ 114

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U)............ 115

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu) .............................. 116

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 117

C-12 General Engineering Laboratories 6-28-99 .......................................................................... 118

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 119

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 120

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 121

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 122

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 123

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 124

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................. 125

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 126

C-21 Body Counter Results 12-8-81 ............................................................................................. 127

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 128

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 129

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 130

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 131

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 132

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 133

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 134

C-29 ABACOS-Plus 3-6-96........................................................................................................... 135

C-30 ABACOS-Plus 11-15-01....................................................................................................... 136

C-31 ABACOS-Plus 6-14-01......................................................................................................... 137
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 107 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-1. Urinalysis Record Card and HSDS – Urinalysis Detail (1) (first activity date on the 

HSDS portion: 10-29-54).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 108 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-2. Urinalysis Record Card and HSDS – Urinalysis Detail (2) (first activity date on the 

HSDS portion 8-19-53).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 109 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-3. Urinalysis Record Card and HSDS – Urinalysis Detail (3) (first activity date on the 

HSDS portion: 1-6-58).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 110 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-4. HSDS – Urinalysis Detail (1) (first activity date 9-17-58).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 111 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-5. HSDS – Urinalysis Detail (2) (first activity date 3-19-73).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 112 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-6. Analytical Report – Bioassay Analysis Data 3-15-93.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 113 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-7. Analytical Report – Bioassay Analysis Data 10-28-93.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 114 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-8. Form 1 – Sample Results 1-29-96.
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 115 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-9. Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U).
Document No. ORAUT-TKBS-0011-5 Revision No. 03 Effective Date: 09/30/2014 Page 116 of 177

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-10. Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu).
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-11. Form 1 – Sample Results – Quanterra, Richland 7-31-98.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-12. General Engineering Laboratories 6-28-99.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-13. Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results).
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-14. Health Physics – Body Counter Information 12-8-65.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-15. Health Physics – Body Counter Information 5-16-68.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-16. Health Physics – Body Counter Information 8-26-68.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-17. Health Physics – Body Counter Information 9-16-70.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-18. Radiation Dosimetry – Body Count Results 10-3-74.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-19. Radiation Dosimetry – Body Count Results 5-30-75.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-20. Radiation Dosimetry – Body Count Results 1-9-78.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-21. Body Counter Results 12-8-81.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-22. Radiation Dose Assessment – Body Count Results 7-22-83.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-23. Radiation Dose Assessment – Body Count Results 5-18-83.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-24. Radiation Dose Assessment – Body Count Results 2-21-84.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-25. Radiation Dose Assessment – Body Count Results 3-22-84.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-26. Radiation Dose Assessment – Body Count Results 10-10-85.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-27. Radiation Dose Assessment – Body Count Results 3-6-89.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-28. Internal Dosimetry – Lung Count Results 11-23-93.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-29. ABACOS-Plus 3-6-96.
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Figure C-30. ABACOS-Plus 11-15-01.
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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-31. ABACOS-Plus 6-14-01.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

TABLE OF CONTENTS

SECTION TITLE PAGE

D.1 Purpose ............................................................................................................................... 142

D.1.1 Data Overview.......................................................................................................... 142

D.1.2 Bioassay Data Selection........................................................................................... 142

D.1.3 Analysis.................................................................................................................... 143

D.2 Intake Modeling.................................................................................................................... 146

D.2.1 Assumptions............................................................................................................. 146

D.2.2 Bioassay Fitting ........................................................................................................ 148

D.2.3 Material Types.......................................................................................................... 148

D.2.3.1 Uranium ..................................................................................................... 148

D.2.3.2 Plutonium................................................................................................... 148

D.3 Assigning Intakes and Doses ............................................................................................... 149

D.3.1 Intake Rate Summary............................................................................................... 149

D.3.2 Dose Assignment ..................................................................................................... 151

D.4 Coworker Data Figures ........................................................................................................ 151

LIST OF TABLES

TABLE TITLE PAGE

D-1 Summary of uranium urinary excretion rate analyses, 1953 to 1988 .................................... 144

D-2 Summary of plutonium urinary excretion rate analyses, 1952 to 1988.................................. 145

D-3 241Am lung count bioassay data for individualized 239Pu Type S fits...................................... 147

D-4 Derived uranium intake rates, 1953 to 2005......................................................................... 150

D-5 Derived Type M plutonium intake rates, 1952 to 2005.......................................................... 150

D-6 Derived Type S plutonium intake rates, 1952 to 2005 .......................................................... 151

D-7 IMBA-derived uranium intake rates ...................................................................................... 161

D-8 IMBA-derived plutonium/americium intake rates, Type M..................................................... 167

D-9 IMBA-derived plutonium/americium intake rates, Type S ..................................................... 177

LIST OF FIGURES

TABLE TITLE PAGE

D-1 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, Type F ................................................................................................................ 151

D-2 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type F. ............................................................................................................... 152
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D-3 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type F..................................................................................... 152

D-4 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 84thpercentile, Type F ................................................................................................................ 153

D-5 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, Type M................................................................................................................ 153

D-6 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type M................................................................................................................ 154

D-7 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type M.................................................................................... 154

D-8 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 84thpercentile, Type M................................................................................................................ 155

D-9 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 50thpercentile, Type S................................................................................................................ 155

D-10 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 50thpercentile, Type S................................................................................................................ 156

D-11 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 50thpercentile, Type S................................................................................................................ 156

D-12 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 50thpercentile, Type S................................................................................................................ 157

D-13 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, Type S................................................................................................................ 157

D-14 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 84thpercentile, Type S................................................................................................................ 158

D-15 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 84thpercentile, Type S................................................................................................................ 158

D-16 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 84thpercentile, Type S................................................................................................................ 159

D-17 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 84thpercentile, Type S................................................................................................................ 159

D-18 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, Type S................................................................................................................ 160
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D-19 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type S .................................................................................... 160

D-20 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/1988, 84th-percentile, Type S .................................................................................... 161

D-21 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, Type M................................................................................................................ 162

D-22 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 50thpercentile, Type M................................................................................................................ 162

D-23 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 50thpercentile, Type M................................................................................................................ 163

D-24 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 50thpercentile, Type M................................................................................................................ 163

D-25 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, Type M................................................................................................................ 164

D-26 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 84thpercentile, Type M................................................................................................................ 164

D-27 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 84thpercentile, Type M................................................................................................................ 165

D-28 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 84thpercentile, Type M................................................................................................................ 165

D-29 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 50-percentile, Type M....................................................................................... 166

D-30 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 84th-percentile, Type M.................................................................................... 166

D-31 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, Type S................................................................................................................ 167

D-32 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 50thpercentile, Type S................................................................................................................168

D-33 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 50thpercentile, Type S................................................................................................................ 168

D-34 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 50thpercentile, Type S................................................................................................................ 169
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D-35 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, Type S................................................................................................................ 169

D-36 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, Type S................................................................................................................ 170

D-37 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 84thpercentile, Type S................................................................................................................ 170

D-38 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 84thpercentile, Type S................................................................................................................ 171

D-39 Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 84th-percentile, Type S .................................................................................... 171

D-40 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, Type S................................................................................................................ 172

D-41 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 50thpercentile, Type S................................................................................................................ 172

D-42 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 84thpercentile, Type S................................................................................................................ 173

D-43 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

50th-percentile, Type S........................................................................................................ 174

D-44 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

50th-percentile, Type S........................................................................................................ 174

D-45 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

50th-percentile, Type S........................................................................................................ 175

D-46 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

84th-percentile, Type S........................................................................................................ 175

D-47 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

84th-percentile, Type S........................................................................................................ 176

D-48 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

84th-percentile, Type S........................................................................................................176
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

D.1 PURPOSE

Some employees at DOE sites were not monitored for internal ionizing radiation exposure, or the 

records of such monitoring are incomplete or unavailable. In such cases, data from monitored 

coworkers can be used to estimate an individual’s possible exposure. The purpose of this attachment 

is to provide monitored coworker information for calculating and assigning occupational internal doses 

to employees at RFP for whom no or insufficient bioassay monitoring records exist.

D.1.1 Data Overview

This section provides information on the general selection characteristics of the data and the methods 

of analysis. More detailed radionuclide-specific information is provided in Section D.2. Data analysis 

for 1989 and later data were performed by NIOSH (2006b).

D.1.2 Bioassay Data Selection

Urinalysis data for uranium and plutonium from 1952 to 1988 were extracted from the Comprehensive 

Epidemiology Data Resource (CEDR) database. There were just over 300,000 records in the 

urinalysis database. Four cases had a date before 1952: one each in 1950 and 1951 and two that 

appeared to be date errors (years incorrectly entered as 1911 and 1923).

The RFP HIS-20 database was obtained after the coworker analysis had been performed. A 

comparison of CEDR and HIS-20 was made. The databases are comparable but provide slightly 

differing results in some cases. These differences sometimes suggest CEDR could be slightly more 

favorable to the claimant while, in other cases, the data suggest HIS-20 could be slightly more 

favorable to the claimant. For the majority of the data, the results are similar. In addition, concern 

was expressed by the Rocky Flats Working Group that the number of samples in HIS-20 and CEDR 

were different in some cases. NIOSH demonstrated that the intakes that were predicted by either 

database were almost identical. However the concern on the part of the Working Group persisted. It 

was suggested and agreed that the use of the 95th percentile internal coworker intake for 

unmonitored workers with nontrivial exposure potential would satisfy this concern. It was also agreed 

that this situation and this policy were specific to Rocky Flats, and would set no precedent to be 

applied elsewhere.

In most cases, both the uranium and plutonium results were recorded as dpm/24 hr. However, the 

DU units are date-dependent. Through April 1964 the units were μg/24 hr; from May 1964 to 1988 the 

units were dpm/24 hr. Micrograms of uranium were converted to dpm by a 0.89 multiplier determined 

from the IMBA isotopic abundances for DU. Once converted to dpm, the uranium data were assumed 

to be entirely 234U (Note: This assumption has no impact the statistics). See Section D.3.1 for 

additional discussion on using 234U for the analyses.

All of the uranium and plutonium urinalysis results were recorded either as positive numbers or zeros. 

In general, a zero entry meant the result was less than some reporting level, but actual results were 

reported after April 6, 1970. Zeros were reported in 176,900 records, a little over half of the results for 

all measurements. Uranium and plutonium urinalysis data with a “1” flag in the “nocalc” field of the 

database (about 2,500 records out of roughly 300,000) should be (and were) excluded from analysis 

because the data did not meet quality objectives.

In vivo 241Am lung data from 1965 to 1988 were extracted from a Microsoft Access table named 

“RFFACW02_RFWB.” There were just fewer than 80,000 241Am records in the lung database. From 
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

1965 through 1971, all results (about 4,000) were reported as zero, with no explanation of what those 

values might have meant. Therefore, no analyses were performed on those data. The 241Am 

activities were quantified only if a known plutonium incident occurred. However, the results were 

sometimes recorded (in counts per minute) when no known incident had occurred. Some results 

were recorded in micrograms or nanocuries. Therefore, careful interpretation of the data units was 

imperative. Positive values began to appear after 1971, but there still were no exclusion instructions 

for when zero values were reported (see the “nocalc” discussion above). Therefore, zero results were 

treated as zeros because no better information was available. Calculations of the lung plutonium 

values that were recorded with the 241Am lung data were determined by using the 241Am data and an 

assumed concentration of 100 ppm (by weight) of 241Am in the plutonium.

In both the urinalysis and lung-counting data sets, badge numbers (the ID column) are associated 

with most records. However, in the urinalysis data, 55,200 records had a “0” in the badge number 

column. It was not determined what a “0” badge ID meant other than, perhaps, to identify unbadged 

personnel. For the urinalysis data, about 34,000 of the “0” badges were plutonium records; 15,000 

were gross alpha (A) and 6,000 were uranium (U). It was decided to treat “0” badge numbers as one 

individual when counting the number of unique individuals in any period. The “sdate” column provided 

the date of each analysis in YYMMDD order.

D.1.3 Analysis

Bioassay data were analyzed by quarter or year, depending on the amount of data available during 

the periods. A lognormal distribution was assumed [131]. As mentioned in the previous section, a 

large fraction of the uranium and plutonium urinalysis data were entered as zeros. In many cases, 

this fact made analysis of the data difficult because so few positive values were reported. Therefore, 

where a reporting level was specified and where zeros were inserted for the actual values in the 

original data (below the reporting level), a linear distribution between zero and the reporting level was 

substituted for the zeros. The linear distribution had the form c/n, 2c/n, 3c/n,..., nc/n where n is the 

number of zero values less than the reporting level c. Using R2 as the fit criterion, this linear 

distribution (alone) fits a lognormal transformation by better than 80% and typically significantly 

improves the goodness of fit for the entire data set. Furthermore, the linear distribution has an 

average equal to half of the reporting value, consistent with the general dose reconstruction practice 

of assigning half of the lower limit of detection for missed dose calculations. As a consequence, 

substituting a linear distribution for these zero entries appears reasonable.

Whenever a linear distribution was substituted for values below a reporting level, the reporting levels 

were used. For EU, these reporting levels were 8.8 dpm/24 hr through 1963, and 20 to 28 dpm/24 hr 

after 1963. For DU, the reporting levels were 5.8 dpm/24 hr through April 1964, 20 to 28 dpm for May 

1964 to 1979, and actual measured values thereafter. For plutonium, these reporting levels were 

0.88 dpm/24 hr through 1961, 0.2 dpm/24 hr for 1962 to April 1970, and actual measured values after 

April 1970. The reporting level for gross alpha through 1963 was 8.8 dpm/24 hr (assigned as EU) and 

0.9 dpm/24 hr thereafter (assigned as plutonium). No reporting level was given for americium-in-lung 

measurements.

After log-transforming the data, the 50th- and 84th-percentile values were determined for each period 

using the method described in ORAUT (2004). Tables D-1, D-2, and D-3 show the statistical analysis 

results for uranium, plutonium, and 241Am, respectively.
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Table D-1. Summary of uranium urinary excretion rate analyses, 1953 to 1988.

a

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

7/1/1953 3.727 10.008 5/15/1964 8.368 23.389

2/15/1954 3.866 10.362 8/15/1964 8.161 22.172

5/15/1954 4.161 11.472 11/15/1964 8.297 23.535

8/15/1954 3.732 10.074 7/1/1965 7.823 20.789

11/15/1954 3.409 9.389 7/1/1966 7.432 18.360

2/15/1955 3.225 9.019 7/1/1967 7.445 18.440

5/15/1955 3.333 9.487 7/1/1968 7.430 18.459

8/15/1955 3.434 9.406 7/1/1969 7.509 18.518

11/15/1955 3.442 9.875 7/1/1970 7.440 18.275

2/15/1956 3.310 9.039 7/1/1971 7.421 18.131

5/15/1956 3.497 9.843 7/1/1972 7.316 18.176

8/15/1956 3.635 10.213 7/1/1973 7.403 18.059

11/15/1956 3.302 9.121 7/1/1974 7.388 18.084

2/15/1957 3.460 9.894 7/1/1975 7.378 18.104

5/15/1957 3.492 10.173 7/1/1976 7.418 18.037

8/15/1957 3.655 10.781 7/1/1977 0.172 0.538

11/15/1957 3.700 10.996 7/1/1978 0.893 2.355

2/15/1958 4.089 12.575 7/1/1979 0.444 2.037

5/15/1958 3.739 10.593 7/1/1980 0.241 1.049

8/15/1958 3.907 11.266 7/1/1981 0.178 1.109

11/15/1958 4.705 14.316 2/15/1982 0.237 1.152

2/15/1959 4.381 13.159 5/15/1982 0.062 0.677

5/15/1959 5.518 17.908 8/15/1982 0.016 0.211

8/15/1959 5.544 16.566 11/15/1982 0.112 0.741

11/15/1959 5.887 19.134 2/15/1983 0.221 1.062

2/15/1960 8.806 33.071 5/15/1983 0.432 1.330

5/15/1960 6.856 22.227 8/15/1983 0.327 1.576

8/15/1960 7.476 24.214 11/15/1983 0.072 0.646

11/15/1960 6.602 23.668 2/15/1984 0.273 1.400

2/15/1961 5.944 20.258 5/15/1984 0.221 1.330

5/15/1961 5.722 18.628 8/15/1984 0.133 0.997

8/15/1961 5.574 18.290 11/15/1984 0.065 0.464

11/15/1961 6.598 22.669 2/15/1985 0.034 0.410

2/15/1962 5.862 20.451 5/15/1985 0.030 0.281

5/15/1962 4.692 15.380 8/15/1985 0.040 0.511

8/15/1962 5.654 16.742 11/15/1985 0.037 0.415

11/15/1962 4.397 13.827 2/15/1986 0.029 0.357

2/15/1963 4.166 13.230 5/15/1986 0.033 0.339

5/15/1963 4.175 13.154 8/15/1986 0.018 0.207

8/15/1963 3.841 12.283 11/15/1986 0.022 0.316

11/15/1963 3.601 11.507 7/1/1987 0.057 0.467

2/15/1964 6.354 18.506 7/1/1988 0.059 0.412

a. All results shown in bold are annual rather than quarterly averages.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table D-2. Summary of plutonium urinary excretion rate analyses, 1952 to 1988.

a

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

7/1/1952 2.514 8.198 11/15/1972 0.028 0.168

7/1/1953 0.716 1.046 2/15/1973 0.024 0.145

7/1/1954 0.575 1.053 5/15/1973 0.033 0.180

7/1/1955 0.469 0.919 8/15/1973 0.067 0.305

7/1/1956 0.615 1.264 11/15/1973 0.061 0.268

7/1/1957 2.610 12.006 2/15/1974 0.060 0.224

2/15/1958 2.173 10.041 5/15/1974 0.049 0.189

5/15/1958 1.037 2.872 8/15/1974 0.033 0.144

8/15/1958 1.295 3.801 11/15/1974 0.016 0.109

11/15/1958 0.919 2.581 2/15/1975 0.021 0.104

2/15/1959 0.709 1.542 5/15/1975 0.019 0.095

5/15/1959 0.942 2.276 8/15/1975 0.022 0.200

8/15/1959 0.945 2.482 11/15/1975 0.015 0.097

11/15/1959 0.560 1.211 2/15/1976 0.016 0.144

2/15/1960 0.614 1.353 5/15/1976 0.021 0.102

5/15/1960 0.596 1.221 8/15/1976 0.015 0.104

8/15/1960 0.453 0.955 11/15/1976 0.043 0.184

11/15/1960 0.573 1.528 2/15/1977 0.083 0.262

2/15/1961 0.728 1.625 5/15/1977 0.092 0.245

5/15/1961 0.691 1.377 8/15/1977 0.072 0.190

8/15/1961 0.754 2.035 11/15/1977 0.062 0.188

11/15/1961 0.656 1.645 2/15/1978 0.095 0.307

2/15/1962 0.337 0.809 5/15/1978 0.060 0.199

5/15/1962 0.326 0.735 8/15/1978 0.056 0.201

8/15/1962 0.271 0.589 11/15/1978 0.033 0.134

11/15/1962 0.220 0.431 2/15/1979 0.062 0.237

2/15/1963 0.250 0.467 5/15/1979 0.013 0.100

5/15/1963 0.248 0.496 8/15/1979 0.013 0.087

8/15/1963 0.238 0.432 11/15/1979 0.029 0.139

11/15/1963 0.252 0.562 2/15/1980 0.017 0.106

2/15/1964 0.296 0.810 5/15/1980 0.017 0.064

5/15/1964 0.249 0.483 8/15/1980 0.013 0.061

8/15/1964 0.379 1.668 11/15/1980 0.004 0.035

11/15/1964 0.334 1.066 2/15/1981 0.006 0.037

2/15/1965 0.283 0.757 5/15/1981b 2.25E-04 0.006

5/15/1965 0.348 1.085 8/15/1981 0.005 0.036

8/15/1965 0.221 0.417 11/15/1981 0.008 0.056

11/15/1965 0.266 0.646 2/15/1982b 1.43E-04 0.007

2/15/1966 0.293 0.821 5/15/1982b 3.11E-04 0.011

5/15/1966 0.237 0.554 8/15/1982b 1.37E-04 0.004

8/15/1966 0.213 0.430 11/15/1982b 2.90E-04 0.006

11/15/1966 0.252 0.625 2/15/1983 0.001 0.017

2/15/1967 0.251 0.622 5/15/1983b 3.99E-04 0.008

5/15/1967 0.240 0.565 8/15/1983 0.002 0.016

8/15/1967 0.199 0.413 11/15/1983 0.004 0.029

11/15/1967 0.236 0.535 2/15/1984 0.008 0.050

2/15/1968 0.228 0.526 5/15/1984 0.053 0.222

5/15/1968 0.205 0.461 8/15/1984 0.011 0.071

8/15/1968 0.252 0.585 11/15/1984 0.054 0.196
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Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

11/15/1968 0.278 0.724 2/15/1985 0.010 0.080

2/15/1969 0.292 0.692 5/15/1985 0.025 0.100

5/15/1969 0.266 0.606 8/15/1985 0.014 0.081

8/15/1969 0.240 0.519 11/15/1985 0.017 0.100

11/15/1969 0.264 0.558 2/15/1986 0.005 0.033

2/15/1970 0.242 0.515 5/15/1986 0.004 0.038

5/15/1970 0.165 0.623 8/15/1986 0.007 0.038

8/15/1970 0.100 0.423 11/15/1986 0.008 0.042

11/15/1970 0.120 0.470 2/15/1987 0.004 0.030

2/15/1971 0.091 0.366 5/15/1987 0.005 0.036

5/15/1971 0.055 0.209 8/15/1987 0.008 0.051

8/15/1971 0.073 0.293 11/15/1987 0.008 0.050

11/15/1971 0.061 0.249 2/15/1988 0.003 0.032

2/15/1972 0.046 0.398 5/15/1988 0.002 0.033

5/15/1972 0.046 0.442 8/15/1988 0.005 0.034

8/15/1972 0.029 0.199 11/15/1988 0.006 0.038

a. All results shown in bold are annual averages rather than quarterly averages. Very large results for: badge 395943 

excluded from 1964-1965; badges 164455 and 184168 excluded from quarter 3, 1971; 164455 and 184169 excluded 

from quarter 4, 1971; badge 184106 excluded from quarter 2, 1976.

b. Results for quarter 2, 1981, all of 1982, and quarter 2, 1983 were not used in calculations because there are too few 

results.

D.2 INTAKE MODELING

This section discusses intake modeling assumptions, intake fitting, and the intake materials (uranium 

and plutonium).

D.2.1 Assumptions

Each result in the intake calculations was assumed to be normally distributed [132]. A uniform 

absolute error of 1 was applied to all results, thus assigning the same weight to each result. IMBA 

requires results to be in units of activity per day; therefore, all urinalysis results were normalized, as 

needed, to 24-hour samples, using 1,400 mL, the volume of urine excreted by Reference Man in a 

24-hour period.

Because of the nature of work at RFP, it is possible that intakes could have been either chronic or 

acute. However, a series of acute intakes can be approximated as a chronic intake. Therefore, 

intakes were assumed to be chronic and were assumed to occur through inhalation, using a default 

breathing rate of 1.2 m3

/hr and a 5-μm AMAD particle size distribution (ICRP 1995).

For intake modeling purposes, all uranium activity was assumed to be 234U. This assumption does not 

affect the fitting of the data for intake determination because all uranium isotopes behave the same 

biokinetically and the isotopes considered in this analysis all have long half-lives in relation to the 

assumed intake period. ICRP Publication 68 dose coefficients (also referred to as dose conversion 

factors) for 234U are 7% to 31% larger than those for 235U, 236U, and 238U (ICRP 1995). Therefore, the 

assumption that the intake is 100% 234U provides a result favorable to the claimant.

For plutonium, 239Pu was assumed for the intake modeling. Before the mid-1970s, plutonium 

urinalysis was performed by chemical separation followed by the counting of all alpha-emitting 
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Table D-3. Americium-241 lung count bioassay data for individualized 239Pu Type S fits.

a

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

Effective 

sample date

50th

percentile

(dpm/24 hr)

84th

percentile

(dpm/24 hr)

7/1/1972 6.73E-05 0.003 5/15/1981 0.016 0.140

2/15/1973 0.005 0.059 8/15/1981 0.016 0.138

5/15/1973 0.010 0.107 11/15/1981 0.016 0.136

8/15/1973 0.025 0.188 2/15/1982 0.013 0.126

11/15/1973 0.009 0.095 5/15/1982 0.011 0.111

2/15/1974 0.005 0.067 8/15/1982 0.010 0.102

5/15/1974 0.007 0.080 11/15/1982 0.009 0.081

8/15/1974 0.007 0.079 2/15/1983 0.006 0.066

11/15/1974 0.007 0.079 5/15/1983 0.002 0.031

2/15/1975 0.017 0.150 8/15/1983 0.005 0.055

5/15/1975 0.027 0.180 11/15/1983 0.008 0.063

8/15/1975 0.039 0.244 2/15/1984 0.005 0.058

11/15/1975 0.048 0.278 5/15/1984 0.006 0.058

2/15/1976 0.043 0.261 8/15/1984 0.005 0.054

5/15/1976 0.044 0.254 11/15/1984 0.008 0.067

8/15/1976 0.017 0.133 2/15/1985 0.004 0.042

11/15/1976 0.012 0.111 5/15/1985 0.005 0.051

2/15/1977 0.010 0.097 8/15/1985 0.003 0.035

5/15/1977 0.008 0.082 11/15/1985 0.003 0.037

8/15/1977 0.007 0.061 2/15/1986 0.004 0.049

11/15/1977 0.004 0.051 5/15/1986 0.007 0.054

2/15/1978 0.008 0.083 8/15/1986 0.005 0.057

5/15/1978 0.007 0.070 11/15/1986 0.004 0.043

8/15/1978 0.007 0.066 2/15/1987 0.008 0.072

11/15/1978 0.004 0.045 5/15/1987 0.005 0.051

7/1/1979 0.012 0.108 8/15/1987 0.009 0.091

2/15/1980 0.026 0.195 11/15/1987 0.009 0.072

5/15/1980 0.020 0.159 2/15/1988 0.006 0.061

8/15/1980 0.021 0.171 5/15/1988 0.008 0.073

11/15/1980 0.027 0.207 8/15/1988 0.005 0.043

2/15/1981 0.018 0.151 11/15/1988 0.004 0.042

a. All results shown in bold are annual averages rather than quarterly averages.

isotopes of plutonium (i.e., 238Pu, 239Pu, and 240Pu). In the mid-1970s, alpha spectroscopy was used 

to differentiate between them. For this modeling, the gross plutonium alpha results are assumed to 

represent only alphas from 239Pu, which results in approximately a 2% overestimate of the modeled 

intakes. This assumption is made to enable consistent modeling of data from both types of urinalysis.

Starting in 1972, lung counts were performed to determine the lung burden of 241Am. These lung 

counts can be used to determine the intake of plutonium. For each plutonium material type, the more 

limiting value of the intakes as determined by the americium lung counts or plutonium urinalysis was 

used. Use of the higher value (from the less sensitive bioassay method for a given material type) 

would be inconsistent with the available bioassay records because a higher intake would result in 

higher-than-observed bioassay results from the more sensitive bioassay method.
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D.2.2 Bioassay Fitting

IMBA was used to fit the bioassay results to a series of inhalation intakes. Data from 1952 through 

1988 were fit as a series of chronic intakes.

The intake assumptions were based on observed patterns in the bioassay data. Periods with 

constant chronic intake rates were chosen by selecting periods where the bioassay results were 

similar. A new chronic intake period was started if the data indicated a significant sustained change in 

the bioassay results. By this method, 1952 through 1988 was divided into multiple chronic intake 

periods.

D.2.3 Material Types

See Section 5.2 for source term solubilities.

D.2.3.1 Uranium

Because the uranium isotopes at RFP have very long radiological half-lives and the material is 

retained in the body for long periods, excretion results are not independent. For example, an intake in 

the 1950s could contribute to urinary excretion in the 1980s and later. To avoid potential 

underestimation of intakes for people who worked at RFP for relatively short periods, each chronic 

intake was fit independently, using only the bioassay results from the single intake period for Type S 

solubility. For Type M and F solubility, this approach was used where it was determined that earlier 

intake rates significantly biased later intake rates. This method results in a potential overestimate of 

intakes for exposures extending through multiple assumed intake periods. Only the results within the 

intake period were selected for use in fitting each period. Excluded results are shown in light gray in 

the figures at the end of this attachment.

Uranium urinalysis results were analyzed with IMBA to derive intake rates for 1953 to 1988. Excretion 

data are shown in Table D-1. The solid lines in Figures D-1 and D-2 show the individual fits to the 

50th-percentile excretion rates for type F material. Figure D-3 is the combined fit for all the intake 

periods. Figure D-4 shows the overall fit to the 84th-percentile excretion rates for type F material. 

The same intake periods were applied for both percentiles because the values followed a similar 

pattern. Similarly, Figures D-5 and D-6 show the individual fits to the 50th-percentile excretion rates 

for type M material. Figure D-7 is the combined fit for all the intake periods. Figure D-8 shows the 

overall fit to the 84th-percentile excretion rates for type M material. Figures D-9 to D-13 and D-14 to 

D-18 show the individual fits to the 50th-and 84th- percentile excretion rates for type S material, 

respectively. Figures D-19 and D-20 show the 50th- and 84th-percentile predicted excretion rates, 

respectively, from all type S intakes. Table D-7 tabulates the derived intake rates for Types F, M, and 

S materials at both the 50th- and 84th-percentile levels along with the associated geometric standard 

deviations (GSDs). Data for 1989 and later were from NIOSH (2006b).

D.2.3.2 Plutonium

Plutonium urinalysis results were analyzed with IMBA using type M and S materials to derive intake 

rates for 1952 to 2005. As with Type S uranium, plutonium isotopes at RFP have very long 

radiological half-lives and the material is retained in the body for long periods, so excretion results are 

not independent. To avoid potential underestimation of intakes for people who worked at RFP for 

relatively short periods, each chronic intake was fit independently, using only the bioassay results 

from the single intake period for both Type M and S solubility. This method results in a potential 
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

overestimate of intakes for exposures extending through multiple assumed intake periods. Only the 

results within the intake period were selected for use in fitting each period. Excluded results are 

shown in light gray in the figures. Tables D-2 and D-3 provide the bioassay data that were used to 

perform the fits.

Plutonium Type M—The solid lines in Figures D-21 to D-24 and D-25 to D-28 show the individual fits 

to the 50th- and 84th-percentile excretion rates for type M materials, respectively. The same intake 

periods were applied for both percentiles because the values followed a similar pattern. Figures D-29 

and D-30 show the 50th- and 84th-percentile predicted excretion rates, respectively, from all type M 

intakes. In addition, intake rates for Type M plutonium based on lung-counting measurements of the 

associated americium-241 were also derived. The plutonium urinalysis results were determined to be 

more limiting and thus were used for the final values. Table D-8 lists the 50th- and 84th-percentile 

intake rates along with the associated GSD determined from plutonium urinalysis. For comparison, 

the intake rate determined from the americium lung counts at the 50th percentile level are also given. 

Data for 1989 and later were from NIOSH (2006b).

Plutonium Type S— The solid lines in Figures D-31 to D-35 and D-36 to D-40 show the individual fits 

to the 50th- and 84th-percentile excretion rates for type S materials, respectively. The same intake 

periods were applied for both percentiles because the values followed a similar pattern. Figures D-41 

and D-42 show the 50th- and 84th-percentile predicted excretion rates, respectively, from all type S 

intakes. Figures D-43 to D-45 and D-46 to D-48 show the individual fits to the 50th- and 84thpercentile 241Am lung count data. Table D-9 lists the 50th- and 84th-percentile intake rates along with 

the associated GSD determined from the plutonium urinalysis and 241Am lung count data. Data for 

1989 and later were from NIOSH (2006b).

D.3 ASSIGNING INTAKES AND DOSES

This section describes the derived intake rates and provides guidance for assigning doses. For each 

intake period discussed below, the 50th- and 84th-percentile calculated intakes were used to 

determine the GSD of the data. The GSD along with the geometric mean were used to calculate the 

95th-percentile intake rate. Data for 1989 and later were from NIOSH (2006b). In 1993, the 

Secretary of Energy formally announced the end of nuclear production at Rocky Flats. Remediation 

was completed at the RFP in late 2005. Coworker intakes should be assigned, when applicable, up 

through 2005. Only environmental intakes should be assigned after 2005.

D.3.1 Intake Rate Summary

Multiple intake periods were fit to the derived 50th- and 84th-percentile uranium excretion data. 

Table D-4 summarizes the 95th-percentile uranium intake rates derived from the fits.

Similarly, multiple intake periods were fit to the derived 50th- and 84th-percentile plutonium excretion 

and americium lung burden data for Type M material and Type S material. Table D-5 summarizes the 

95th-percentile plutonium intake rates derived from the fits for Type M material.

For types S and Super S material, Table D-6 provides the urinalysis and lung-count based intakes 

rates to be used.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table D-4. Derived uranium intake rates, 1953 to 2005.

Period

95th percentile (dpm/d)

Type F 

material

Type M 

material

Type S 

material

1953–1958 74.2 303 5,266

1959 130 763 20,322

1960 212 763 20,322

1961 135 502 11,600

1962 163 502 11,600

1963 84.7 502 11,600

1964 161 516 7,391

1965–1976 118 516 7,391

1977–1988 8.52 11.5 458

1989–1993 5.21 21.8 426

1994–2005 1.64 6.72 101

Table D-5. Derived Type M plutonium intake rates, 1952 

to 2005.

Period

95th percentile (dpm/d) 

Type M material

1952–1961 718

1962–1969 190

1970–1979 75.6

1980–1988 26.7

1989–1993 47.6

1994–2005 2.21

The Table D-6 intake rates should be used as follows:

• For doses to systemic organs, use the systemic intake rates in Table D-6 in accordance with 

the guidance in ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly Retained in the 

Lung (ORAUT 2010).

- Type Super S coworker doses to systemic organs should be calculated using 

urinalysis-based intakes for all periods.

• Doses to the lungs and thoracic lymph nodes, gastrointestinal tract, and extrathoracic regions 

(nonsystemic organs) should be based on the nonsystemic intake rates in Table D-6 in 

accordance with the guidance in ORAUT-OTIB-0049 (ORAUT 2010).

– Type Super S coworker doses to nonsystemic organs should be calculated based on 

the lung count-based intakes for 1972 through 1988.

– Type Super S coworker doses to nonsystemic organs should be calculated based on 

the urinalysis-based intakes for 1952 through 1971 and for 1989 through 2005.

• For all coworker type Super S adjustments, the date of the last bioassay sample should be 

assumed to be equivalent to the intake period end date in Table D-6. Note that type Super S 

adjustments should be made separately for each intake period in Table D-6.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table D-6. Derived Type S plutonium intake rates, 1952 to 2005.

Systemic intake rates Nonsystemic intake rates

Intake period 

start date

Intake period 

end date

intake rate, 

dpm/d 239Pu

Intake period 

start date

Intake period 

end date

intake rate, 

dpm/d 239Pu

1/1/1952 12/31/1961 11,243 a 1/1/1952 12/31/1961 11,243 a

1/1/1962 12/31/1969 3,368 a 1/1/1962 12/31/1969 3,368 a

1/1/1970 12/31/1979 1,168 a 1/1/1970 12/31/1971 1,168 a

1/1/1980 12/31/1993 385 a 1/1/1972 12/31/1976 953 b

1/1/1994 12/31/2005 34.2 a 1/1/1977 12/31/1982 863 b

1/1/1983 12/31/1988 419 b

1/1/1989 12/31/1993 385 a

1/1/1994 12/31/2005 34.2 a

a. Urinalysis-based intake rates.

b. Lung count-based intake rates.

D.3.2 Dose Assignment

Doses to be assigned to individuals are calculated from the 95th-percentile intake rates [133]. Dose 

reconstructors should select the material type that is the most favorable to the claimant.

The constant distribution is selected in IREP, with the calculated dose entered as Parameter 1.

D.4 COWORKER DATA FIGURES

Figure D-1. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1976, 50thpercentile, Type F.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-2. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 50thpercentile, Type F.

Figure D-3. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type F.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-4. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1988, 84thpercentile, Type F.

Figure D-5. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1976, 50thpercentile, Type M.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-6. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 50thpercentile, Type M

.

Figure D-7. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type M.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-8. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1988, 84thpercentile, Type M.

Figure D-9. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1958, 50thpercentile, Type S.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-10. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1959 to 12/31/1960, 50thpercentile, Type S.

Figure D-11. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1961 to 12/31/1963, 50thpercentile, Type S.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-12. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1964 to 12/31/1976, 50thpercentile, Type S.

Figure D-13. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 50thpercentile, Type S.
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INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-14. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1958, 84thpercentile, Type S.

Figure D-15. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1959 to 12/31/1960, 84thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-16. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1961 to 12/31/1963, 84thpercentile, Type S.

Figure D-17. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1964 to 12/31/1976, 84thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-18. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 84thpercentile, Type S.

Figure D-19. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/1988, 50th-percentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-20. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/1988, 84th-percentile, Type S.

Table D-7. IMBA-derived uranium intake rates (dpm/day).

Years

Type F Type M Type S

50% 84% GSD 50% 84% GSD 50% 84% GSD

1953–1958 13.37 37.91 2.84 54.75 154.8 2.83 936.9 2,676 2.86

1959 19.7 61.99 3.15 102.7 347.5 3.38 2,768 9,300 3.36

1960 27.23 94.74 3.48 102.7 347.5 3.38 2,768 9,300 3.36

1961 21.62 65.97 3.05 71.85 234.2 3.26 1,680 5,438 3.24

1962 16.27 65.97 4.05 71.85 234.2 3.26 1,680 5,438 3.24

1963 16.27 44.36 2.73 71.85 234.2 3.26 1,680 5,438 3.24

1964 27.26 80.39 2.95 112.8 284.2 2.52 1,630 4,086 2.51

1965–1976 27.26 66.26 2.43 112.8 284.2 2.52 1,630 4,086 2.51

1977–1988 0.597 3.004 5.03 2.443 6.263 2.56 28.6 154.3 5.40
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-21. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 50thpercentile, Type M.

Figure D-22. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1969, 50thpercentile, Type M.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-23. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1970 to 12/31/1979, 50thpercentile, Type M.

Figure D-24. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1988, 50thpercentile, Type M.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-25. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 84thpercentile, Type M.

Figure D-26. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1969, 84thpercentile, Type M.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-27. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1970 to 12/31/1979, 84thpercentile, Type M.

Figure D-28. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1988, 84thpercentile, Type M.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-29. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), from all intakes 1/1/1952 to 

12/31/1988, 50-percentile, Type M.

Figure D-30. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), from all intakes 1/1/1952 to 

12/31/1988, 84th-percentile, Type M.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table D-8. IMBA-derived plutonium/americium intake rates, Type M.

Year

Plutonium urinalysisbased results, dpm/dayAmericium lung count-based 

results, 50th percentile

Pu 50% Pu 84% GSD Am, pCi/day Pu, dpm/day

1952–1961 121 357.3 2.95

1962–1969 43.5 106.5 2.45

1970–1971 7.05 29.82 4.23

1972–1976 7.05 29.82 4.23 0.387 178

1977–1979 7.05 29.82 4.23 0.280 129

1980–1982 1.622 8.907 5.49 0.280 129

1983–1988 1.622 8.907 5.49 0.124 57.3

Figure D-31. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 50thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-32. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 50thpercentile, Type S.

Figure D-33. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 50thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-34. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 50thpercentile, Type S.

Figure D-35. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 50thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-36. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 84thpercentile, Type S.

Figure D-37. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 84thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-38. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 84thpercentile, Type S.

Figure D-39. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 84thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-40. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 84thpercentile, Type S.

Figure D-41. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 50thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-42. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 84thpercentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-43. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

50th-percentile, Type S.

Figure D-44. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

50th-percentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-45. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

50th-percentile, Type S.

Figure D-46. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

84th-percentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Figure D-47. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

84th-percentile, Type S.

Figure D-48. Predicted americium bioassay results (line) calculated using IMBA-derived americium

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

84th-percentile, Type S.
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ATTACHMENT D

INTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table D-9. IMBA-derived plutonium/americium intake rates, Type S.

Year

Plutonium urinalysisbased results, dpm/dayAmericium lung count-based 

results, 50th percentile, pCi/day

Pu 50% Pu 84% GSD Am 50% Am 84% GSD

1952–1961 1925 5628 2.92

1962–1969 781.1 1899 2.43

1970–1971 112 465.8 4.16

1972–1976 112 465.8 4.16 0.0862 0.595 6.91

1977–1979 112 465.8 4.16 0.0534 0.465 8.70

1980–1982 13.69 104.1 7.60 0.0534 0.465 8.70

1983–1988 13.69 104.1 7.60 0.024 0.0240 9.12

1989–1993 13.69 104.1 7.60

1994–2005 0.7993 7.838 9.81